ML23062A419

8 to Updated Final Safety Analysis Report, Appendix C, Structural Design Criteria. (Redacted)

ML23062A419
Person / Time
Site:	Peach Bottom
Issue date:	03/03/2023
From:	Exelon Generation Co
To:	Office of Nuclear Reactor Regulation
References
Download: ML23062A419 (1)

Text

PBAPS UFSAR APPENDIX C - STRUCTURAL DESIGN CRITERIA C.1 CLASSIFICATION OF STRUCTURES C.1.1 General Certain station structures must remain functional and/or protect vital equipment and systems, both during and following the most severe natural phenomenon which is postulated to occur at the site. In order to establish the loadings and loading combinations for which each individual structure is to be designed, buildings and their structural systems are separated into the following two seismic classes with respect to a seismic design requirements.

Seismic Class I - Seismic Class I structures and equipment are those whose failure could increase the severity of the design basis accident, and cause release of radioactivity in excess of 10 CFR 50.67 limits, or those essential for safe shutdown and removal of decay heat following a LOCA.

Seismic Class II - Seismic Class II structures and equipment are those whose failure would not result in the release of significant radioactivity and would not prevent reactor shutdown. The failure of seismic Class II structures may interrupt power generation.

A structure designated seismic Class II shall not degrade the integrity of any structure designated seismic Class I. Although a structure, as a whole, may be seismic Class I, less essential portions may be considered seismic Class II if they are not associated with loss of function, and their failure does not render the seismic Class I portions inoperable.

Seismic Class II structures are structurally separated from seismic Class I structures by means of expansion joints to provide for unequal deflections associated with independent movements of the structures. The arrangement is such that in the unlikely event that a Class II structure should collapse, it would not impair the safety function of the Class I structure.

The criteria for the relative movements under maximum earthquake loadings require that the clearance provided exceeds the combined movements. The relative movements under these loadings are accommodated by expansion joints at adjoining structures and by built-in flexibility for piping systems. A dynamic analysis has shown that the cumulative maximum displacements of adjoining concrete structures will be about one-half of the clearance provided.

APPENDIX C C.1-1 REV. 28, APRIL 2021

PBAPS UFSAR In the case of structures defined as partially Class I and partially Class II rigidly interconnected, the Class I portion is checked to assure it can carry any loads that may be transmitted from the connected Class II structure.

The following list itemizes the structures, equipment, and process systems which fall under the two seismic classes defined above.

C.1.2 Seismic Class I Structures and Systems Class I Structures Drywell, vents, torus, and penetrations Reactor building Spent fuel pool Reactor vessel support pedestal Main control room complex (including cable spreading room, emergency switchgear rooms, and battery rooms)

Radwaste building Diesel generator building Pump structure (portion containing critical service water pumps)

Emergency heat sink facility, including cooling tower Stack Structures required to protect seismic Class I equipment CADS liquid N2 tank building Recombiner building Class I Equipment and Systems Nuclear steam supply systems:

Reactor vessel and internals, including:

CRD housing CRD guide tube CRD CRD cap screw Control rod CRD thermal sleeve and key In-core housing Feedwater sparger Jet pump adapter Shroud Top guide Core support Core support and top guide aligner APPENDIX C C.1-2 REV. 28, APRIL 2021

PBAPS UFSAR Core plate stud Jet pump riser brace Jet pump assembly Jet pump instrument penetration seal Differential pressure and liquid control line Core spray line and clamp Head cooling spray nozzle (for Unit 2 only)

Dry tube Power range monitor installation hardware Power range detector Orificed fuel support Fuel channel Fuel assembly Reactor vessel supports and stabilizers Control rod drive system (equipment required for scram operation)

Control rod drive housing supports Recirculating piping, including valves and pumps Main steam piping out to second isolation valve Nuclear boiler system safety valves Nuclear boiler system relief valves Piping connections from the reactor vessel, up to and including the first isolation valve external to the drywell Core standby cooling systems (CSCS)

Standby liquid control system (except for the test tank and test connections)

High pressure service water system Emergency service water system Standby gas treatment system Fuel storage facilities, to include spent fuel and new fuel storage racks Reactor building crane Circulating water pump structure crane (designed to Seismic Class I requirements)

Standby power systems:

Station batteries (except balance-of-plant battery and 24 volt neutron monitoring batteries)

Standby diesel generators Emergency buses and other electrical gear for onsite power supply to engineered safeguards and nuclear safety systems Instrumentation and controls:

Reactor level instrumentation Reactor manual control system Control rod instrumentation (portions)

CADS APPENDIX C C.1-3 REV. 28, APRIL 2021

PBAPS UFSAR C.1.3 Seismic Class II Structures and Systems Class II Structures Turbine building Shop and warehouse Administration building Water treatment building Pump structure, except for portion affecting critical service water systems Intake screen structure Cooling towers and cooling tower pump structures for circulating water Off-gas filter station Auxiliary boiler house Guardhouse Outdoor electrical switchgear structures Sewage treatment plant Radwaste onsite storage facility Recirc ASD Structure Class II Equipment and Systems Turbine-generator system and transformers Condensers Turbine building crane Feedwater heaters and pumps Condensate storage tanks and pumps Refueling water storage tank Station auxiliary power buses Offsite AC power system Radwaste systems Reactor water cleanup system Condensate filter-demineralizer system Compressed air system Reactor building cooling water system Turbine building cooling water system Instrument N2 system All other piping and equipment not listed under seismic Class I 24 volt neutron monitoring batteries Feedwater zinc injection system Hydrogen Water Chemistry System APPENDIX C C.1-4 REV. 28, APRIL 2021

PBAPS UFSAR C.2 STRUCTURAL DESIGN BASIS Structures are designed for dead loads, live loads, seismic loads, and wind loads in accordance with applicable codes and as described in the following paragraphs. The loading conditions, and combinations thereof, are determined by the function of the structure and its importance in meeting the station safety and power generation objectives.

C.2.1 Dead and Live Loads The structures in the power plant complex are designed for the dead loads and live loads to which the structures will be subjected. Roofs of all the structures are designed for a snow load of 30 psf.

C.2.2 Seismic Loads The design of seismic Class I structures is based on a dynamic analysis using the spectrum response curves developed for the site. The design of seismic Class I equipment is based on a dynamic analysis using either acceleration spectrum response curves or acceleration time histories developed at points of attachment, the method of analysis being dependent on the nature of the equipment.

The list of Class I (seismic design) structures, equipment, and systems is presented in paragraph C.1.2. All structures listed in this table as Class I structures were seismically analyzed by the response spectra method, except the portion of the pump structure containing critical service water pumps was seismically analyzed by the time-history method.

The structures are analyzed for the following magnitudes of ground acceleration:

a. Design earthquake considers a maximum horizontal ground acceleration of 0.05g. Under this condition, stresses due to the earthquake combined with stresses due to other operational loadings are limited to the working stress levels of the materials used in the structures except as noted in paragraph C.2.6.3. The customary increase in normal allowable working stress due to earthquake is not used.

b. Maximum credible earthquake (MCE) considers a horizontal ground acceleration of 0.12g. Under this condition, APPENDIX C C.2-1 REV. 28, APRIL 2021

PBAPS UFSAR stresses due to the earthquake combined with stresses due to other operational loadings are allowed to approach the yield strength of the materials and are limited to 90 percent of yield stress (fy) for the steel and 85 percent of ultimate compressive stress (f'c) for the concrete. In addition, all items required for safe shutdown will not lose their function.

Proof of design adequacy is accomplished by showing the criteria stated for steel and concrete for the MCE condition are not exceeded and thus the structures comply with the definition of seismic Class I in paragraph C.1.1. Structural deformations and deflections calculated are well within the linear-elastic range and cause only low stresses.

c. Vertical ground accelerations associated with the design earthquake and MCE are 67 percent of the corresponding horizontal acceleration spectrums; namely, 0.033g for the design earthquake and 0.08g for the MCE.

Table C.2.1 shows the damping factors which are used for excitations associated with the design earthquake and the MCE.

Vertical seismic stresses are not severe because they represent only a fractional increase in the dead load which the structure carries. Since the frequencies of the modes associated with vertical motion are normally large, it is sufficient to design the vertical elements for the maximum vertical ground acceleration without a detailed dynamic analysis of the structure.

The reactor building is nearly symmetrical about both perpendicular axes. The lack of symmetry is not sufficient to significantly alter stresses and may be safely ignored. However, to account for so-called "accidental" torsion, after evaluating the worst cases an arbitrary conservative allowance of 20 percent was made on all forces.

Parametric studies were carried out to determine the relative influence of the numerous variables involved which verified the adequacy of the assumption.

The vertical seismic response can be divided into two categories.

The first category is the general building motion involving primarily the column or wall elements and the second category is the local response of various beam and slab elements oriented parallel to the ground.

APPENDIX C C.2-2 REV. 28, APRIL 2021

PBAPS UFSAR In general, for a building founded on a rigid foundation the building response will be small compared to the dead load since the building frequencies will be higher than the primary frequencies of the earthquake spectrum.

The beams, slab, equipment, and systems may respond differently than the overall building since their frequencies may correspond to the primary frequencies of the earthquake spectrum.

All Class I equipment and structural elements including columns, walls, beams, and slabs are analyzed and designed to resist the vertical seismic forces together with any other loads as defined in the design criteria. Beams and floors are analyzed to determine their maximum response and frequency. The equipment and systems are designed to resist any amplified beam and floor accelerations.

The seismic Class II radwaste on-site storage facility structure is designed for seismic loadings corresponding to the maximum ground acceleration of 0.05g selected for the Operating Basis Earthquake (OBE). A model analysis using a lumped mass model of the facility was performed using the criteria and methodology described in USNRC Regulatory Guide 1.143. American Concrete Institute standard ACI 318-77, "Building Code Requirements for Reinforced Concrete" was used in the design of the concrete structures. For steel structure design, American Institute of Steel Construction "Specification for the Design, Fabrication and Erection of Structural Steel for Buildings," November 1978 was used. The one-third allowable stress increase was included for steel structures for load combinations involving earthquakes or wind loads. The building foundation is discussed in UFSAR section 2.7.6.4.

Analysis of other seismic Class II structures is based on the design criteria established for the structures in Zone I of the seismic zones as defined by the Uniform Building Code, 1967 Edition.

Class II structures, such as the turbine building, which adjoin Class I structures are arranged and designed in such a way that the possible failure of the Class II building will not endanger the function of any Class I building or system.

Additionally, in the case of the 1997 re-analysis of the Recirculation system piping and the Residual Heat Removal and Reactor Water Clean-up piping inside primary containment for Peach Bottom NCR 97-02267, the seismic analysis was based on NRC APPENDIX C C.2-3 REV. 28, APRIL 2021

PBAPS UFSAR Regulatory Guide 1.60 (Design Response Spectra for Seismic Design of Nuclear Power Plants) with modal combination and spatial components in accordance with Reg. Guide 1.92 (Combining Modal Responses and Spatial Components in Seismic Response Analysis).

These regulatory guides were used because they are required by Regulatory Guide 1.84 when using ANSI Code Case N-411-1 (Alternative Damping Values for Response Spectra Analysis of Class 1, 2, and 3 Piping Section III, Division 1).

C.2.3 Wind Loads The methods used in determining the wind pressures for the radwaste on-site storage facility are in accordance with ANSI A58.1-1972, "Building Code Requirements for Minimum Design Loads in Buildings and other Structures". The storage facility structures are designed to withstand a maximum windspeed of 90 miles per hour. The wind is assumed to occur 30 feet above ground and has a 100-year mean recurrance interval.

The wind loads used in the design of other portions of this plant are derived from Paper 3269, entitled "Wind Forces on Structures,"

published by the American Society of Civil Engineers, Transactions, Volume 126, Part II, 1961, as applied to the Peach Bottom site. The total wind pressures, listed in Table C.2.2, include positive and negative pressures and gust factors.

C.2.4 Tornado Loads The plant is designed to withstand a tornado and remain in a safe shutdown condition to prevent undue risk to the health and safety of the public. Tornado winds traversing the site could damage the reactor building superstructure, turbine building, condensate storage tanks, stack, and incoming power lines. However, the ability to shut down the reactor, the integrity of primary containment, and the capability of heat removal systems necessary to maintain a safe shutdown condition would not be impaired.

The joint occurrence of a tornado with other low probability events (design basis accidents, seismic event, flood, etc.) has a sufficiently small likelihood of occurring that it is not considered in the design of the plant. As with other natural phenomena and special events (as described in Sections 2.4.3.5, 2.5.3.5 and Table 1.4.2), the single failure criterion does not apply to tornado events. As a result, redundant trains or components are generally not necessary to shut down and maintain the plant in a safe shutdown condition following and recovering from a tornado.

APPENDIX C C.2-4 REV. 28, APRIL 2021

PBAPS UFSAR Components required to shut down the plant and maintain it in a safe shutdown condition are located in reinforced concrete structures, underground, or are otherwise protected from the effects of tornados. These components include the following:

Reactor Coolant Pressure Boundary Control Rod Drive System (portion essential for SCRAM)

Systems or portions of systems needed to:

a) Maintain adequate core cooling (e.g., HPCI, RCIC or LPCI) b) Remove decay heat Primary containment and isolation valves Necessary support systems (e.g., HPSW, ESW, intake structure)

Station batteries needed to support required systems Standby diesel-generators and associated switchgear Controls and instrumentation for above systems Main control room Where tornado missile barriers are necessary to protect the integrity of primary containment, the reactor pressure boundary, or equipment necessary to shut down the reactor and maintain it in a safe shutdown condition, failure could affect the operation and function of the primary containment, the reactor primary system, or other safeguards equipment, the following tornado effects are considered in the design of these barriers:

1. External wind forces resulting from a tornado having a horizontal peripheral tangential velocity of 300 mph maximum, which includes the tangential and translational components.

2. Differential pressure of 3 psi between inside and outside of fully enclosed areas. Blowout panels are APPENDIX C C.2-5 REV. 28, APRIL 2021

PBAPS UFSAR included where necessary in the design of the structure to limit pressure differentials.

3. Horizontal missiles (no vertical velocity component) equivalent to a 4 in thick x 12 in wide x 12 ft long wood plank traveling end-on at 300 mph; or a 4,000-lb passenger auto flying horizontally through the air at 50 mph, at not more than 25 ft above ground, with a contact area of 20 sq ft.

At the time of the original design and licensing of the plant, design criteria for non-horizontal missiles (i.e., missiles with a vertical velocity component) did not exist and the site is not committed to any specific criteria for vertical missiles. Changes to the plant design will provide a similar level of protection as existed in the original licensed design.

4. A torsional moment resulting from applying the wind specified in item 1 acting on one-half the length of a building.

Walls of all open compartments were designed to withstand the differential pressure which occurs during the tornado depressurization. Blowout panels are provided to relieve excess positive pressure in all essential parts of the structure.

Building structures housing safeguards equipment are designed to withstand a tornado-induced depressurization rate of 1 psi/sec for 3 sec. To accomplish this objective, all such compartments that are sufficiently leaktight to develop a differential pressure with the outside environment are designed to withstand a differential pressure of 3 psi.

Equipment or structures not required to safely shut down the reactor and maintain it in a safe shutdown condition are not designed to be protected against tornado effects.

Seismic Class I equipment and/or structures either protected by a tornado resistant structure, or whose loss of function during a tornado would not violate the safety requirements of the plant, are not designed against tornado effects.

The structural steel frame of the reactor building upper superstructure is designed to withstand the force of a 300-mph wind without exceeding the yield stress. The reactor building siding and roof decking, however, is designed for the normal wind APPENDIX C C.2-6 REV. 28, APRIL 2021

PBAPS UFSAR loading. When this design wind loading is appreciably exceeded, portions of the siding and decking are expected to be lost.

Connectors for the siding are designed to fail at stress levels associated with tornado loading to assure that the siding will blow away. However, to ensure an adequate load carrying capacity of the structural members, the individual members were designed to take the full load of the tornado if the siding directly affecting that member remained intact. However, the reinforced concrete structure of the reactor building protects the contained equipment necessary for the safe shutdown of the reactor, the primary containment, and the essential heat removal equipment from the effects of a tornado. Tornado effects on the spent fuel pool are discussed in General Electric Topical Report APED-5696. On the sides, the fuel pool is protected against low trajectory missiles by thick concrete walls between the turbine and the pool.

C.2.5 Special Loadings The structures housing critical equipment required for safe shutdown of the plant are designed for special loadings.

C.2.5.1 Turbine Missiles The turbine missile probability will be maintained to less than 1 x 10-5 per year, and the probability of damaging a critical target will be maintained less than 1 x 10-7. This is consistent with Sections 3.5.1.3 and 2.2.3 of the Standard Review Plan.

Section 11.2.4 includes the basis for determining probabilities and the inspection program that has been instituted to maintain the probability of turbine missile generation within acceptable limits.

Missiles from the RCIC turbine were also investigated to assure that they would not damage any critical piping in the vicinity of the turbine. The possibility of this type of missile is very remote.

C.2.5.2 Tornado-Generated Missiles Tornado-generated missiles are discussed in paragraph C.2.4. The concrete shield plug above the drywell is capable of resisting missiles generated in a tornado. The large equipment openings of the diesel generator building have missile-proof doors. The personnel access doors are shielded from such missiles by baffle walls. This concept is typically used throughout the project to protect large openings against effects of tornado winds, depressurization, and tornado-generated horizontal missiles.

APPENDIX C C.2-7 REV. 28, APRIL 2021

PBAPS UFSAR C.2.5.2.1 Tornado Missile Risk Evaluator The Tornado Missile Risk Evaluator (TMRE) is a risk-based methodology developed by the Nuclear Energy Institute (NEI) to assess the risk posed by tornado generated missiles. The methodology is applied to equipment that does not conform to the original licensing basis for protection against tornado missiles. The original licensing basis, as described in Sections C.2.4 and C.2.5.2, relies on physical barriers (typically concrete walls or underground) to provide protection against tornado missiles. Equipment has been identified that is not protected by physical barriers. For this equipment, TMRE has been applied to demonstrate that the risk posed by tornado missiles is sufficiently low to justify not providing physical protection.

The TMRE process determines the likelihood that exposed equipment will be struck and damaged by a tornado missile, using a broad spectrum of missiles (size, weight, speed and direction). The site-specific Probabilistic Risk Assessment (PRA) model is then used to determine the risk (i.e., change in core damage frequency) associated with the exposed equipment.

The risk acceptance criteria contained in NRC Regulatory Guide 1.174 are used to determine if physical protection is warranted.

The TMRE process is only used for equipment that does not conform to the original licensing basis. It is not approved by the NRC for use with new modifications. New modifications must meet design requirements contained in the original licensing basis for protection against tornado generated missiles.

The TMRE methodology in use at the site is described in NEI 17-02, Rev. 1B, with the restrictions and limitations contained in the NRC's approval for the use of TMRE at Grand Gulf Nuclear Station (NRC letter dated June 18, 2019, ML19123A014). The site-specific results of the methodology are contained in Reference 38.

C.2.5.3 Temperature Loads For each seismic Class I structure, temperature loads considered to be significant were included in the design. For example, the biological shield was designed for the normal operating loads listed in Table C.4.5; the reactor pressure vessel pedestal was designed for the loading conditions listed in Table C.4.4; the primary containment shell was designed for the accident conditions listed in Tables M.3.5 and M.3.6; the fuel pool walls were APPENDIX C C.2-8 REV. 28, APRIL 2021

PBAPS UFSAR originally designed for normal allowable stresses. A check under loss-of-fuel pool coolant (i.e., boiling water) indicated that stresses would be still below normal allowable limits. Refer to Section C.2.6.3 for reevaluation of the spent fuel pool.

Higher temperatures than LOCA condition were not considered for other than process equipment normally encountering higher temperatures; however, the stress levels are sufficiently low to be able to tolerate a short duration increase in temperature to 305F and still be within the allowable limits.

Transient stresses do not significantly affect concrete stresses.

However, transients were considered at the point of embedment of the drywell shell. The design basis temperature was initially 281°F during a LOCA. The current bounding drywell temperature, however, occurs during a break of a steam line. A spectrum of steam line break sizes have been evaluated to ensure a bounding drywell temperature profile is established. The most limiting drywell temperature from this analysis is 340°F. Although the drywell environment may see temperatures as high as 340°F for 20 minutes, the most limiting temperature for the drywell shell has been analyzed to be within the design temperature of 281°F (Reference 24).

C.2.5.4 Flood Loads and Flood Protection Structures required for safe shutdown of Units 2 and 3 in the event of the probable maximum flood (PMF), (causing an estimated wave runup Security Related Information Withheld under 10 CFR 2.390 assuming no accident occurs concurrently, are:

Reactor building Main control room complex Diesel generator building Pump structure (portion containing critical service water pumps)

Emergency heat sink facility, including cooling tower Components required for safe shutdown of Units 2 and 3 are:

Reactor vessel and internals CRDS (portion essential for scram)

Recirculation piping system RCICS RHRS High pressure service water system APPENDIX C C.2-9 REV. 28, APRIL 2021

PBAPS UFSAR Emergency cooling system Emergency service water system Standby power systems Instrumentation and controls:

Reactor level instrumentation Reactor pressure instrumentation For description of wave runup superimposed on the PMF refer to subsection 2.4.

For drawings of structures and components listed above see Figures 12.1.1, 12.2.1, and Drawings C-84, and M-2 through M-7. The emergency heat sink structure is shown in Figure C.2.1.

Watertight doors are provided at all structures; waterproofing is installed Security Related Information Withheld under 10 CFR 2.390 and any penetration in the exterior walls is sealed to ensure leaktightness necessary to plant safety.

The integrity of the waterproofing on the external surfaces of vertical walls below grade cannot be checked since such surfaces are inaccessible. Accessible joints are visually inspected and caulked as required on a periodic basis as part of regular plant maintenance.

Plastic waterstops are used at all construction joints to maintain the integrity of joints. Penetrations and conduits in exterior walls are sealed with approved, pre-tested seal details and material which assure leaktightness against ground or flood water.

Penetration seals are installed in accordance with approved specifications and procedures and are inspected to assure proper installation.

C.2.6 Loading Combinations The following paragraphs describe the loading combinations used for the design of the seismic Class I structures. Loads and loading combinations for Class II structures are in accordance with the Uniform Building Code and normal design practice for power plants. Loading combinations used for the design of the primary containment are discussed in Appendix M.

D = Dead load of structure and equipment plus any other permanent loads contributing stress, such as soil or hydrostatic loads, operating pressures, and live loads expected to be present when the plant is operating.

50 psf is considered normal operating live load.

APPENDIX C C.2-10 REV. 28, APRIL 2021

PBAPS UFSAR W = Design wind loading conditions.

W' = Loads due to tornado.

R = Jet force on structure due to rupture of any one pipe.

H = Force on structure due to thermal expansion of pipes under operating conditions. The effect of this loading was considered on individual members where required.

E = Design earthquake load.

E' = MCE load.

T = Temperature load.

F = Flood loading ( Security Related Information Withheld under 10 CFR 2.390 ).

For Class I structures, code allowable stress values are modified since structures of this class must sustain much more severe loads and be more accurately proportioned than structures normally considered under building codes. However, the same codes will still furnish guidance.

The criteria for seismic Class I structures with respect to stress levels and load combinations for the postulated events are noted in the following paragraph.

C.2.6.1 Reactor Building and All Other Seismic Class I Structures

1. D+E Normal allowable code stresses (AISC for structural steel, ACI for rein-forced concrete). The customary increase in normal design stresses, when earthquake loads are considered, is not permitted.

2. D+E' Maximum allowable stresses are as follows:

Steel - 0.9 Fy (yield strength of steel);

Concrete - 0.85 f'c (compressive strength of concrete);

APPENDIX C C.2-11 REV. 28, APRIL 2021

PBAPS UFSAR Reinforcement - 0.9 fy (yield strength of reinforcement).

3. D+W Maximum allowable working stresses may be increased one-third above normal code allowable stresses.

4. D+W' Maximum allowable stresses are as follows:

Steel - 0.9 Fy; Concrete - 0.85 f'c; Reinforcement - 0.9 fy.

5. D+E+T Normal allowable code stresses.

The customary increase in normal design stresses when earthquake is considered is not permitted.

6. D+E'+T Maximum allowable stresses are as follows:

Steel - 0.9 Fy; Concrete - 0.85 f'c; Reinforcement - 0.9 fy.

7. D+F Maximum allowable stresses are as follows:

Steel - 0.9 Fy; Concrete - 0.85 f'c; Reinforcement - 0.9 fy.

C.2.6.2 Reactor Vessel Pedestal

1. D+T+E Normal allowable code stresses (AISC for structural steel, ACI for reinforced concrete). The customary increase in normal design stresses, when earthquake loads are considered, is not permitted.

2. D+T+R Maximum allowable stresses are as follows:

Steel - 0.9 Fy; Concrete - 0.85 f'c; Reinforcement - 0.9 fy.

3. D+T+E' Maximum allowable stresses are as follows:

APPENDIX C C.2-12 REV. 28, APRIL 2021

PBAPS UFSAR Steel - 0.9 Fy; Concrete - 0.85 f'c; Reinforcement - 0.9 fy.

C.2.6.3 Spent Fuel Pool The spent fuel pool has been reevaluated structurally for additional loading due to a loaded spent fuel storage cask, the higher capacity control rod blade racks, the high density fuel racks and increased number of fuel elements. This reevaluation was performed in accordance with the applicable codes and standards identified in Section C.2.7.1.

All loading combinations required by USNRC Regulatory Guide 1.142, USNRC Standard Review Plan 3.8.4, ACI and AISC were evaluated.

The number of combinations to be analyzed were reduced by eliminating combinations governed by others. Final governing equations for the spent fuel pool structure are shown below for concrete structures using strength design methods and for structural steel using plastic design methods.

Load Combinations Reinforced Concrete

1. U = 1.4D + 1.4F + 1.7T0

2. U = 1.4D + 1.4F

3. U = 1.4D + 1.4F + 1.7L + 1.9E

4. U = D + F + L +E' + Ta

5. U = D + F + L +E'

6. U = 1.05D + 1.05F + 1.3L + 1.43E + 1.3T0 Structural Steel

7. Y = 1.7D + 1.7F + 1.7L + 1.7E

8. Y = 1.3D + 1.3F + 1.3L + 1.3E + 1.3T0

9. Y = 1.1 (D + F + L + E' + Ta)

Where: L = Live Load T0 = Operating Temperature Ta = Accident Temperature Loading Assumptions:

The dead load includes the weight of the spent fuel racks, stored fuel, spent fuel pool, and the contributing weight of the adjacent floor slabs, roof, and walls.

APPENDIX C C.2-13 REV. 28, APRIL 2021

PBAPS UFSAR The live load includes the roof snow load, the distributed live loads on the adjacent floor slabs, crane loads and a buoyant weight of a loaded spent fuel storage cask.

Hydrostatic loads consist of vertical and lateral water pressures exerted on the spent fuel pool slab and walls, respectively.

Thermal loads are based on the pool water temperatures resulting from a full core discharge under normal operating conditions, and saturation temperatures for accident conditions. In all cases, a conservative Reactor Building indoor ambient temperature of 68F is used. A stress free temperature of 70F is used.

C.2.6.4 Reactor Building Security Related In with Spent Fuel Storage Cask The Security Related Information Withheld under 10 CFR 2.390 reactor building, at the base of the crane hatch, has been reevaluated structurally for a loaded spent fuel storage cask and the cask transporter, in various configurations. The concrete slab was evaluated using ultimate strength design methods, using the load combinations of section C.2.6.3, above. The structural steel was evaluated using allowable stress design methods, using the load combinations listed in Section C.2.6.1.

C.2.7 Governing Codes and Regulations The design of all structures and facilities conforms to the applicable general codes or specifications listed below except where specifically stated otherwise; for example items 2 and 3.

Each structure was analyzed by methods appropriate for its configuration; this furnished a measure of the stresses the structure would experience under the postulated conditions.

Referenced codes were used as guides to establish reasonable allowable stresses.

1. Uniform Building Code (UBC). 1967 Edition.

2. American Institute of Steel Construction (AISC),

"Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings," Sixth Edition and Ninth Edition (See Note 1).

3. American Concrete Institute (ACI), "Building Code Requirements for Reinforced Concrete," (ACI 318-63) Code Requirements for Nuclear Safety Related Concrete Structures (ACI 349-01) (See Note 2), and "Code APPENDIX C C.2-14 REV. 28, APRIL 2021

PBAPS UFSAR Requirements for Reinforced Concrete Chimneys," ACI 307 (1969).

4. American Welding Society (AWS), "Standard Code for Arc and Gas Welding in Building Construction," (AWS-D.1.0).

5. American Petroleum Institute (API), "Specification No. 650 for Welded Steel Storage Tanks."

6. ASME Boiler and Pressure Vessel Code, "Section III, Class B (governs the design and fabrication of the drywell and suppression chamber), 1965 Edition, with applicable addenda published to April, 1967.

7. U.S. Army Corp of Engineers (Regulations with respect to dredging and construction).

8. American Society of Civil Engineers Paper No. 3269, "Wind Forces of Structures."

9. American Iron and Steel Institute (AISI), "Specification for the Design of Light Gage Cold-formed Steel Structural Members."

10. Commonwealth of Pennsylvania Department of Labor and Industry "Building Regulations for Fire and Panic."

11. Electric Power Research Institute (EPRI) "Visual Weld Acceptance Criteria," EPRI Report No. NP-5380 Volume 1:

Visual Weld Acceptance Criteria for Structural Welding at Nuclear Power Plants (NCIG-01, Revision 2), September 1987 See Notes 1 and 2 below.

C.2.7.1 Spent Fuel Pool Reevaluation The spent fuel pool has been evaluated structurally for additional loading due to a loaded spent fuel storage cask, the higher capacity control rod blade racks, the increased number of fuel elements and high density fuel storage racks in accordance with the following codes and standards:

1. American Concrete Institute (ACI), "Building Code Requirements for Reinforced Concrete," (ACI 318-83) and "Code Requirements for Nuclear Safety Related Structures,"

(ACI 349-80)

APPENDIX C C.2-15 REV. 28, APRIL 2021

PBAPS UFSAR

2. American Institute of Steel Construction (AISC),

"Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings," 1978

3. U.S. Nuclear Regulatory Commission, "Standard Review Plan 3.8.4, 'Other Seismic Category I Structures,'" Revision 1, NUREG-0800, July 1981

4. U.S. Nuclear Regulatory Commission, letter from B.R.

Grimes to All Power Reactor Licensees, April 14, 1978, with enclosure entitled "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications," including Supplement, dated January 18, 1979 Note 1: AISC 9th Edition may be used for evaluations that are not addressed in the 6th Edition.

Note 2: NRC Regulatory Guide 1.199 approves the use of ACI 349-01, Appendix B for concrete anchorage evaluations.

APPENDIX C C.2-16 REV. 28, APRIL 2021

PBAPS UFSAR TABLE C.2.1 DAMPING FACTORS Percent of Critical Damping Maximum Design Credible Earthquake Earthquake Reinforced concrete structures 2.0 5.0 Steel framed structures 2.0 5.0 Welded steel assemblies 1.0 2.0 Bolted and riveted assemblies 2.0 5.0 Seismic Class I piping systems

• 0.5 0.5 1.0 (Unit 3 only)

• 1997 Re-analysis of the Recirculation system piping, and the Residual Heat Removal and Reactor Water Clean-up piping inside primary containment for Peach Bottom NCR 97-02267 and ASME Code Case N-411-1 as shown below:

APPENDIX C C.2-17 REV. 21, APRIL 2007

PBAPS UFSAR APPENDIX C C.2-18 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.2.2 WIND LOADS Pressure (q) - (psf)

Class I Class II Structures - Structures-100-Yr 50-Yr Height-Feet Recurrence Recurrence 0-50 25 20 50-150 35 25 150-400 45 30 Over 400 55 40 APPENDIX C C.2-19 REV. 21, APRIL 2007

PBAPS UFSAR C.3 ANALYSIS OF CLASS I STRUCTURES C.3.1 Scope The loads, loading combinations, and allowable limits described in this appendix apply only to seismic Class I structures and components. The criteria in this appendix are intended to supplement applicable industry design codes where necessary to provide design safety margins which are appropriate to extremely reliable structures and components when account is taken of rare events associated with an MCE or postulated LOCA or a combination thereof.

Seismic Class I components are not always designed by application of the criteria using analytical techniques. Rather, the design of some components may be based upon test results, empirical evidence, or by comparison with similar items.

The seismic Class I concrete and steel structures are designed considering three inter-related primary functions for the design loading combinations described in paragraph C.2.6. The first consideration is to provide structural strength equal to or greater than that required to sustain the combination of design loads and provide protection to other seismic Class I structures and components. Design code allowable stresses appropriate for the elastic design techniques were used as a guide for all stress limitations under normal design conditions. Higher stresses approaching yield for steel and ultimate for concrete were permitted under the MCE and similar conditions and as noted in paragraph C.2.6.3. Typical stresses under various conditions have been tabulated in Tables C.4.1 through C.4.5, and when these are compared with ultimate strengths, safety factors are readily apparent. The second consideration is to maintain structural deformations within such limits that seismic Class I components and/or systems will not experience a loss of function.

Deformations experienced by structures under the loss of function criteria were checked and found, by elastic analysis, to be of such a small magnitude as to assure the structure would function as required. Typical deformations of the reactor building are shown in Figures C.3.8 and C.3.9. The third consideration is to limit excessive containment leakage by preventing excessive deformation and cracking where containment integrity is required.

Structural design and construction were performed in such a way as to prevent concrete cracking insofar as possible by mix design, pour limitation, and curing precautions. The stress limits in the code should result in very limited cracking on the order of a few APPENDIX C C.3-1 REV. 25, APRIL 2015

PBAPS UFSAR hundredths of an inch. Such cracking would not significantly affect the leak resistance of the structure.

C.3.2 Structural Analysis In general, the structural analysis is performed utilizing the "Working Stress Design" method as defined in "ACI Standard Building Code Requirements for Reinforced Concrete" (ACI 318-63),

and in the AISC Manual of Steel Construction (Sixth Edition).

"Finite element stress analysis" and other techniques are also used where applicable or necessary.

Load combinations and allowable limits on stresses are as shown in paragraph C.2.6. The maximum permissible calculated concrete compression is limited to 0.85 f'c (design compressive strength of concrete) and the maximum permissible calculated concrete shear is as given in ACI 318-63, Chapter 17, for loading involving R and E'.

Concrete structures designed for no loss of function criteria have been proportioned so as not to exceed 0.9 fy tension in the reinforcing steel and 0.85 f'c compression in the concrete. For bending, stresses have been determined on a straight line stress distribution assumption. This yields maximum allowable moments less than the ultimate strength moment as calculated by ACI-318-63 Code Section 1601. For bending every section is "under reinforced" so that the reinforcing steel reaches its allowable stress before the concrete, thus assuring ductility and reserve strength against structural collapse.

For both reinforcing steel and concrete the design criteria is:

normal allowable stresses were not increased when considering operating loads with design earthquake loads. (Spent fuel pool reevaluation used ultimate strength method for the design earthquake - see C.2.6.3.) No loss of function criterion as listed in paragraph C.2.2 was used for MCE, tornado loads, flood loads, or pipe rupture jet loads when combined with normal loads.

Bond and anchorage for reinforcing steel is treated as required by ACI 318-63.

There are no loading conditions such as pressure which would cause net tension across a section resulting in biaxial and triaxial tension when combined with other loads, and thus reduce the shear strength, bond, and anchorage strength of reinforcing bars.

However, there are loading conditions which produce biaxial stresses on certain members, similar to that experienced by a two-APPENDIX C C.3-2 REV. 25, APRIL 2015

PBAPS UFSAR way slab. This condition is covered by ACI code allowable stresses which were used in the design except for no loss of function criteria loadings. For these criteria, reinforcing bar lap lengths and anchorage lengths that were used to develop the bars for their maximum code allowable stress are adequate to develop the higher stresses produced.

The allowable shear stresses for the no loss of function criteria are presented in Tables C.4.1, C.4.2, and C.4.4 Structural steel members designed for failure criteria have been proportioned so as not to exceed 0.9 Fy in bending and tension, 0.5 Fy in shear, and 1.5 Fa as defined in the AISC-63 code, subsection 1.5.1. Thus, the minimum factors of safety become 1.11 for bending and tension, 1.15 for shear, and from 1.11 to 1.28 for axial compression.

C.3.3 Seismic Analysis of Structures The method used in the seismic analysis consists of the following four steps:

1. Formulation of the mathematical model of the structure or structures to be analyzed.

2. Determination of natural frequencies and mode shapes.

3. Finding the acceleration (g) levels from the response spectra curves.

4. Determination of the response of the structure to the earthquake in terms of moments, shears, and displacements.

The mathematical model of the structure consists of lumped masses and stiffness coefficients. At appropriate locations within the building, points are chosen to lump the weights of the structure.

Between these locations, properties are calculated for moments of inertia, cross-sectional areas, and effective shear areas. The properties of the model are utilized in a computer program, applying unit loads at the mass points to obtain the flexibility coefficients of the building.

The natural frequencies and mode shapes of the structures are obtained by a Bechtel computer program, CE617. The program utilizes the flexibility coefficients and lumped weights of the modes. The flexibility coefficients are formulated into a matrix APPENDIX C C.3-3 REV. 25, APRIL 2015

PBAPS UFSAR and inverted to form a stiffness matrix. The program then uses the technique of diagonalization by successive rotations to obtain the natural frequencies and mode shapes. Appropriate damping values of individual materials are presented in Table C.2.1.

The basic description of the earthquake is provided by spectrum response curves. Separate curves are used for the design earthquake of 0.05g horizontal acceleration and the MCE of 0.12g horizontal acceleration. These curves are presented in Figures C.3.1 and C.3.2. Additionally, 1997 re-analysis of the Recirculation system piping, and the Residual Heat Removal and Reactor Water Clean-up piping inside primary containment for Peach Bottom NCR 97-02267 used Figure C.3.1a and C.3.2a as required by NRC Regulatory Guide 1.60 (Design Response Spectra for Seismic Design of Nuclear Power Plants). This regulatory guide was used because it is required by Regulatory Guide 1.84 when using ASME Code Case N-411-1 (Alternative Damping Values for Response Spectra Analysis of Class 1, 2, and 3 Piping Section III, Division 1). The response of the structure to the earthquake is obtained by using the spectrum response technique. Appropriate acceleration levels are read from the earthquake spectrum curve corresponding to the natural frequencies of the structure. The mode shapes and lumped weights are utilized to calculate an effective weight associated with each mode.

These effective weights and the spectrum curve acceleration levels are utilized to obtain an effective force for each mode. Then, the mode shapes are used again to distribute the effective modal forces of each mode throughout the structure in order to obtain forces at each point for each mode. These forces, on a modal basis, are used as separate loading conditions to obtain the response of the structure. The individual response values per mode at different points for shear moments and displacements are combined on an absolute basis. All mode shapes of the structural system which have natural frequencies below 30 Hz are used or a minimum of four modes.

The response spectrum specified for the site design earthquake and response spectrum generated from acceleration time-history record of the July 12, 1952, Taft, California S69E Earthquake normalized for the 5 percent design earthquake are compared in Figure C.3.12 for 2 percent of critical damping except for main steam line Piping Inside Containment, since only this was used for developing floor spectrum curves. For evaluation of main steam line inside containment, the response spectrum specified for the site design earthquake and response spectrum generated from acceleration time-history records of the July 12, 1952, Taft, California S69E APPENDIX C C.3-4 REV. 25, APRIL 2015

PBAPS UFSAR Earthquake normalized for the 5 percent design earthquake are compared in Figure C.3.12J (Reference 26) for 2 percent of critical damping since only this was used for developing DE floor spectrum curves.

For MCE loading for main steam line inside containment, the response spectrum specified for the site design earthquake and response spectrum generated from acceleration time-history record of the July 12, 1952, Taft, California S69E Earthquake normalized for the 12 percent maximum credible earthquake are compared for 2 and 5 percent damping in Figures C.3.12K and C.3.12L (Reference

25) since only these were used for developing MCE floor spectrum curves. The response spectrum for the 1997 Re-analysis of the Recirculation system piping and Residual Heat Removal and Reactor Water Clean-up piping inside primary containment for Peach Bottom NCR 97-02267 is compared to the site design earthquake in Figure C.3.12A.

To obtain floor spectrum curves for the MCE, the values obtained from the 2 percent damping design earthquake are multiplied by 2.4 (0.12/0.05) except for main steam line piping inside containment.

Since the higher damping for the MCE is thus not considered, values employed are very conservative. For analysis of the main steam line inside containment, the MCE floor spectra curves were obtained using the structural MCE damping value specified in Table C.2.1 instead of multiplying design earthquake by 2.4 (Reference 26).

The time-history technique is used to develop spectrum curves at selected points on the structure for use in equipment analysis.

Since some of the points from the time-history spectrum fall below the site response spectrum, the ratio of the accelerations obtained by the spectrum response technique to the accelerations from the time-history analysis was used as a multiplying factor to increase the time-history spectrum for the Class I structures as appropriate.

Figure C.3.3 shows the mathematical model used for the seismic analysis of the coupled system of the reactor building, reactor vessel pedestal with sacrificial shield, and the reactor vessel.

The model of the reactor vessel used in this coupled system was approximate and was used to study its effect on the reactor building. Figure C.3.3A shows the mathematical model used to generate response spectra curves for the 1997 re-analysis of the Recirculation system piping, and the Residual Heat Removal and Reactor Water Clean-up piping inside primary containment for Peach APPENDIX C C.3-5 REV. 25, APRIL 2015

PBAPS UFSAR Bottom NCR 97-02267. The seismic model in Figure C.3.3A is reconstituted (Reference 25) and was used to develop the spectra for the Main Steam analysis inside containment (Reference 26).

The seismic analysis of the reactor vessel and its internals is discussed in subsection C.5, "Components."

The seismic moments and shears obtained from the analysis were used for the structural design of the buildings with particular emphasis on the seismic overturning, connections of the members, and arrangement of the reinforcing in the concrete. Figures C.3.4 through C.3.11 show moments, shears, displacements, and accelerations for the reactor building.

These graphs represent the values of moments and shear used in the structural design of buildings. These values were checked from time to time to evaluate the effects of the changes associated with the design development of the project, and to assure that the design values used were always conservative.

To assure the aseismic integrity of equipment, an earthquake time-history is selected whose raw spectrum response curve is greater than or equal to the site design spectrum response curve.

This time-history is applied at the base of the building to generate, at selected elevations, additional time-histories and spectrum response curves. These time-histories and spectrum response curves are then utilized to assure the aseismic integrity of the equipment. Other seismic Class I structures were also dynamically analyzed following the same procedure.

APPENDIX C C.3-6 REV. 25, APRIL 2015

PBAPS UFSAR C.4 IMPLEMENTATION OF STRUCTURAL CRITERIA This subsection illustrates the loads and load combinations (subsection C.2) and structural static and dynamic analysis (subsection C.3) used in the structural design of seismic Class I structures.

The design analysis of the primary containment is presented in Appendix M, "Containment Report." Loads, load combinations, and methods of analysis used for the design of the primary containment are described in detail in Appendix M. That appendix also summarizes the actual stresses in the containment vessel under various loading conditions.

This subsection briefly discusses typical structural elements of the reactor building and summarizes the actual stresses in these elements in Tables C.4.1 through C.4.5.

Design procedures used for the reactor building were also used for the other seismic Class I structures, such as the diesel-generator building, the radwaste building, and the pump structure. Design procedures are identical; stresses in various elements of these structures are not illustrated.

C.4.1 Reactor Building Floor System The selection of a particular floor system, precast concrete, poured in place concrete, composite construction, or steel and metal deck, was based on an evaluation of the economics, construction schedule and sequence, shielding requirements, and structural requirements. The reactor building has more than one type of system.

Allowable stress design methods are typically utilized to evaluate floor systems. In a few cases, the newer method of ultimate strength design has been utilized to reevaluate new loading configurations. These cases are described in Section C.2.6.

Although the following example has been reevaluated using the ultimate strength design method, the paragraphs below and associated Table C.4.1 describe the original evaluation method as it demonstrates the primary method used throughout the Class I areas of the plant.

Because of its critical function and particularly heavy loading, the floor system for the ground level operating floor Security Related Information Withheld is selected to illustrate the implementation of design criteria.

APPENDIX C C.4-1 REV. 27, APRIL 2019

PBAPS UFSAR The beam selected is in the area where the railroad track enters the building. In addition to the usual railcar loading, it also supports the spent fuel cask on a special railcar. This area is also designed for a -1,000 psf live load to accommodate the transfer of heavy equipment, such as the recirculating pump motor, which may have to be transported through the railroad lock. This area was also evaluated for a live load of 125-ton cask with a 70-ton transporter.

The floor is designed to include vertical seismic loading simultaneously with the full design live load.

Wind (W), tornado (W'), and jet loads (R) do not act on this particular portion of the floor and, therefore, are omitted.

Table C.4.1 tabulates the design stresses, allowable stresses, and loading combinations as they apply to the particular beam illustrated. The design stresses are within the allowable stresses and the system is structurally adequate.

C.4.2 Reactor Building Concrete Wall The south wall (column line 8) of the Unit 2 reactor building is selected to illustrate the implementation of the design criteria.

This wall experiences several loading combinations. It is a shear wall for the seismic forces due to an earthquake (E or E') in the east-west direction. It is also a peripheral basement wall for the torus and experiences soil and hydrostatic loads (part of D) on its south face. In the superstructure of the building, this wall is designed to withstand normal wind loads (W), as well as tornado loads (W').

The combination of these loads is critical for the design of the south wall. The governing design condition was D+E.

The wall design was also investigated for the effects of horizontal tornado missiles and the thickness of the wall was determined to be adequate.

Design stresses and allowable stresses are tabulated in Table C.4.2. The design stresses are within the allowable limits.

C.4.3 Reactor Building Superstructure The basic frame of the superstructure consists of stepped crane columns and a 12-ft deep truss.

APPENDIX C C.4-2 REV. 27, APRIL 2019

PBAPS UFSAR The roof, consisting of purlins framing from truss to truss and metal deck, forms a rigid diaphragm. The bottom chords of the trusses are tied together with horizontal bracing.

All the peripheral columns, stepped crane columns on the east and west (column lines B and J), and wind columns on the north and south (column lines 8 and 18) are braced, and support girts in turn support the metal siding. The effect of metal siding as a diaphragm was neglected. Column bracing is designed for wind (W) and earthquake (E) loading. Columns are designed for dead and live loads (D), wind loads (W), and earthquake loads (E). The superstructure is also designed for tornado loads (W') on the assumption that all or part of the metal siding is blown away due to the tornado, and the basic superstructure frame is subjected to full tornado winds of 300 mph. Under this condition, the frame will withstand the loading without failure. The stresses may exceed normal allowable stresses, but will not exceed 90 percent of the yield stress of structural steel. The trusses are designed on the same basis. The effect of suction on the metal deck was taken into account.

In the design of the reactor building's superstructure, as well as its floors and walls, concentrated loads are structurally accommodated by the addition of special restraining systems as for jet loads and the support members were designed to carry the reactions. Member proportions were established to provide adequate protection wherever there was a probability of missile impingement.

Table C.4.3 summarizes the loading combinations, method of analysis, resulting design stresses, and allowable stresses. The design stresses are well within the limits of the allowable stresses.

C.4.4 Reactor Pedestal The reactor pedestal was investigated for various loads: dead and live load (D), earthquake (E or E'), temperature (T) associated with an accident condition for a thermal gradient of 70F, and jet forces (R) associated with a pipe rupture. Jet forces on pipe restraints attached to the sacrificial shield and pedestal were also investigated. The overall design was based on very conservative assumptions to allow for the complex interactions of the various loads.

APPENDIX C C.4-3 REV. 27, APRIL 2019

PBAPS UFSAR Incorporated into the design of the reactor shield is the capability to withstand, without failure, the internal pressure and coincident jet impingement loads resulting from failures of high-pressure lines in the shield space region (from the outside diameter of the reactor shield to the outside diameter of the reactor vessel). Failure of the reactor vessel (including nozzles) is not considered credible; however, the consequences of safe-end failures are given full consideration. Safe-ends, even though attached by the reactor vessel manufacturer, are not considered to be an integral part of the reactor vessel but are regarded as a transition piece between the reactor vessel and the primary piping. Although steps have been taken, in light of recent experience, to effectively preclude safe-end failures, the design criteria developed for the reactor shield considers a full spectrum of breaks up to a double-ended recirculation line break at the nozzle to safe-end weld.

The design of the reactor pedestal included additional base anchorage above that required by calculations to assure no adverse affects from secondary stresses.

The loading combinations, resulting stresses, and allowable stresses are tabulated in Table C.4.4. The design stresses are within the allowable stresses resulting in a high safety factor.

Stresses are shown at the baseSecurity Related Information Withheld under 10 CFR 2.390 and at an intermediate level, Security Related Information Withheld under 10 CFR 2.390 . Base stresses are the maximum. Stresses decrease from this point to those shown Security Re

. Security Related Information Withheld under 10 CFR 2.390 Basic reinforcement is uniform throughout the pedestal and is the basis for the reported stresses.

For Unit 2, primarily for construction considerations, permanent steel plates were used in lieu of wood forms. In the design of the anchorage of the base of the pedestal, the steel liner plates were anchored to provide the added anchorage at the base and studs were provided to secure the plate to the concrete. Based on economic considerations developed from the Unit 2 experience, it was decided to remove one or both of the liner plates for Unit 3 and to replace the anchorage so deleted by an additional row of dowels near the pedestal wall centerline.

With the added dowels, the Unit 3 reactor pressure vessel pedestal meets the same design requirements as the Unit 2 pedestal. As constructed, an inside liner similar to that for Unit 2 was used for Unit 3, and the outside of the Unit 3 pedestal was formed.

With regard to the anchorage requirements, no credit was taken for APPENDIX C C.4-4 REV. 27, APRIL 2019

PBAPS UFSAR the inner liner plate. The removal of the outside liner did not decrease the pedestal's capability to resist all postulated loads.

Temperature effects were accounted for in the design of the reactor pedestal using the ACI 505-54 method of analysis for the steady-state condition. Since the thickness of concrete is large, the time required to form a higher gradient than that used is beyond the expected time of exposure and therefore not considered critical.

For justification of allowable stresses see paragraph C.3.2.

The ring girder is designed to transfer the vertical and horizontal loads of the reactor pressure vessel skirt flange to the top of the reactor pressure vessel support pedestal.

The horizontal shears on the reactor pressure vessel skirt flange are transferred to the top flange of the ring girder by 60 A490 high strength bolts in the same friction-type connection as is designed in the AISC Code.

The amount of frictional force available to resist horizontal shear is directly proportional to the normal pressure (proof load) between the reactor pressure vessel skirt flange and top flange of the ring girder. The total frictional force and the coefficient of sliding friction is independent of the areas in contact, so long as the total pressure remains the same. The friction-type connection of the reactor pressure vessel skirt flange to the ring girder, in which some of the bolts lose a part of their clamping force (proof load) due to applied tension during an earthquake, suffer no overall loss of frictional shear resistance. The bolt tension produced by the moment is coupled with a compensating compressive force on the other side of the axis of bending.

The total frictional force due to a coefficient of friction of 0.15 and a proof load of 405 kips per bolt is 3,650 kips or 2.8 times the design basis earthquake shear load of 1,300 kips or 1.4 times the MCE shear load of 2,600 kips. However, if the coefficient of friction is assumed zero, the bolts acting as bearing-type connections could resist a total horizontal shear of 4.15 (at AISC Code stresses) times the design basis earthquake shear load of 1,300 kips or 2.6 (at 90 percent yield stresses) times the MCE shear load of 2,600 kips. Therefore, the high strength bolt connections of the reactor pressure vessel skirt flange to the top flange of the ring girder, with or without friction, are more than adequate for the respective design load.

APPENDIX C C.4-5 REV. 27, APRIL 2019

PBAPS UFSAR The vertical loads on the reactor pressure vessel skirt flange are transferred to the top of the reactor pressure vessel support pedestal by the ring girder as a bearing plate. The ring girder is designed according to AISC Code.

For stresses between the pedestal and the spherically shaped base see Table C.4.4 Security Related Information Withheld under 10 CFR 2.390 .

Moderate friction (0.2) on the shear ring connecting the spherically shaped concrete base to the steel drywell will prevent translation. However, bearing on the external concrete structure will also prevent translation and, therefore, no reliance on the friction factor is necessary. Table C.4.4 reflects this.

C.4.5 Drywell Shielding Concrete The drywell shielding concrete structure, due to its irregular shape and loading combinations, was designed by the finite element method and checked by other methods. Large openings were accounted for in the analysis by introducing lower stiffness elements at regions where they occur in the finite element analysis. In the design check, a shell approach was used and the worst resulting stresses were incorporated in the design. The openings were designed as frames to carry stresses around discontinuities. The most conservative results were used in the final design.

Table C.4.5 illustrates the design stresses under various loading combinations consisting of dead and live loads (D), temperature (T) with a thermal gradient of 70F, and earthquake (E or E').

The design gradient of 70 is based on a heat balance analysis and a drywell bulk average temperature of 145F. From Table C.4.5 stresses are quite low, providing sufficient margin to accommodate higher gradients. In any case, the time it takes to heat up this large mass of concrete is very long and, therefore, the accident transients do not significantly affect the concrete stresses. The design values tabulated are based on the finite element method used for the structural design. The structure was assumed as axisymmetrical with allowance made for local discontinuities at penetrations. In the finite element program, an uncracked section of concrete was assumed, yielding very conservative results.

The drywell shielding concrete is not subjected to tornado loads (W') or wind loads (W). The indirect application of jet forces (R) was investigated as a special case. Reactions from jet forces were taken into consideration in the design of the biological shield concrete. Load combinations listed in Table C.4.5 are APPENDIX C C.4-6 REV. 27, APRIL 2019

PBAPS UFSAR representative of typical sections of structure, and since reactions at piping and equipment anchor points constitute localized conditions, they were not listed. The concrete is capable of withstanding the jet forces, as a localized load, should the drywell yield locally without rupture and close the 2-in air gap between the drywell and shield. The 2-in air space around the drywell is open at all penetrations through the biological shield and drained by pipes at the bottom. Since the 2-in gap is ventilated at several places to the atmosphere, it cannot become pressurized due to temperature inside the drywell.

The effects due to thermal expansion of pipes under operating condition (H) are insignificant inside the primary containment when compared to pressure loadings and, therefore, are not included. No hot pipes are rigidly attached to the drywell shell.

Expansion bellows have been provided at critical hot penetrations.

H represents either the operating or accident condition, whichever is greater. Normal allowable stresses were not increased when using this loading.

The design stresses under all loading combinations are within the allowable limits.

C.4.6 The Sacrificial Shield The sacrificial shield was designed without considering the concrete for any structural purpose, except the lower 10 ft of the wall. The forces considered were: seismic forces, pipe loading, pipe restraints, platform loads, and jet load reaction. The 27-in thick cylindrical structure consists of 12 steel columns equally spaced and continually tied by a 1/4-in thick steel plate on the inside and outside of the columns. For seismic design the sacrificial shield was modeled as a beam on a spring support at the top and fixed at the bottom. A space truss model was also used to check individual sections subject to combination of stresses with the aid of a computer. After the integrity of the overall structure was ascertained, local stresses, connections, and discontinuities were investigated. Proper account was made of nonaxisymmetric loads.

The design allowable stresses are based on the AISC Specifications for the Design, Fabrication, and Erection of Structural Steel for Buildings, without any increase in the normal allowable stresses when design earthquake loads are included. For the MCE condition APPENDIX C C.4-7 REV. 27, APRIL 2019

PBAPS UFSAR in combination with jet loadings, the stresses are allowed to reach 90 percent of yield stresses.

Reactor vessel penetrations which penetrate the sacrificial shield are closed with removable shield plugs which fit around the penetration pipe. The removable shield plugs allow access for in-service inspections.

C.4.7 Main Steam Pipe Chase The main steam pipe chase is designed to sustain a static pressure of 10 psig based on the vent area available and steam release from a single pipe failure. A panel designed to blow out at less than 1.25 psig is provided to eliminate the possibility of a higher pressure buildup.

The design and construction is according to ACI 318-63 and uses the same criteria as for Class I structures under a no-loss-of-function criterion.

The design allowable stresses (safety factors) are based on the ACI 318-63 working stress method. Since the blowout panel constitutes a portion of the boundary of the secondary containment, it must remain intact for those situations for which secondary containment integrity is required. The panel was analyzed for its ability to withstand without failure the effects of the MCE (0.12g) and the design was found to be adequate.

Analyses were made of the resultant stresses in the piping within the steam tunnel for the MCE to assure that those lines with a potential for energy release to the tunnel sufficient to cause panel failure will meet the stress requirements of seismic Class I piping as defined in Appendix A.

C.4.8 Spent Fuel Pool The fuel pool, together with the dryer-separator storage pool, forms a channel-shaped beam supported in the middle by the biological concrete shield structure and at the outer ends by the building walls.

In order to minimize the possibility of pool leakage, the pools are lined with stainless steel.

A finite element analysis was performed to determine the maximum allowable fuel rack loads which can be imposed on the pool slab.

The analysis included the effects of the water in the pool, including the fluctuation of pressure due to seismic acceleration APPENDIX C C.4-8 REV. 27, APRIL 2019

PBAPS UFSAR and sloshing. Thermal effects due to normal operating and accident conditions were also included.

The available section strengths for reinforced concrete elements are calculated by the strength design method in accordance with ACI 318 and ACI 349. Axial force/moment and axial force/shear interaction diagrams are generated for the entire spent fuel pool structure. These interaction diagrams were then used to manually check each critical section. The axial force/shear interaction diagram for the spent fuel pool floor includes the transverse shear strength of the steel beams. The available section strengths for structural steel members for axial loads plus bending are determined by plastic design methods in accordance with AISC.

The section strengths required to carry the fuel pool loading are based on results from the finite element analyses. Required section strengths in terms of shear forces and bending moments are determined for each element in the spent fuel pool structure and for each of the governing load combinations.

C.4.9 Concrete Block Walls The typical block wall design was based on a static coefficient of acceleration of 0.2g. When the plant design was completed, walls were checked dynamically based on their location in the structure to ensure the block walls could withstand the worst load combinations associated with the MCE and verify that the preliminary design was conservative. A technical report in response to NRC IE Bulletin 80-11 was prepared and issued on April 17, 1981, which describes in greater detail the re-evaluation criteria for the block walls. The conclusion of this re-evaluation is noted therein and necessary modifications have been completed to assure that the block walls do not prejudice the integrity of any Class I equipment.

C.4.10 Strength Tests and Crack Control for Concrete The entire Unit 2 reactor building and radwaste building concrete, as well as most of the remainder of the concrete in place at the time of the Bechtel Corporation report, "Concrete Strength Survey Report," dated June, 1970, were surveyed by the Swiss hammer method. Reports are included in the referenced report. Lewis H.

Tuthill, past president of ACI, reviewed the number of cores taken and stated that non-destructive testing was better than drilling the structure full of holes with possible impairment of strength APPENDIX C C.4-9 REV. 27, APRIL 2019

PBAPS UFSAR and agreed the Swiss hammer readings gave uniform results of the concrete in place.

Twenty-two thousand (22,000) cu yd of concrete was placed and 117 representative cores taken or approximately 1 core for every 188 cu yd of concrete placed. The project specification states one set of six cylinders for every 150 cu yd of concrete placed. This gives a ratio of approximately (1:7) 1 core to 7 cylinders, and when the Swiss hammer tests are included with the cores, the ratio of cores to cylinders is greater than one, and the structure was determined to be adequate for its intended function.

ACI 318 paragraph 504(a) states one test will be taken for every 150 cu yd. Two specimens shall be made for each test or 20,000/150 = 133 tests or 266 specimens all to be broken at 28 days. However, ACI 301 paragraph 1704(c) states that at least 3 cores be taken in areas of concrete that were considered deficient. Only five areas were deficient. Therefore, at least 15 cores are required to satisfy ACI 301. With the Swiss hammer tests, in addition to the cores, the structure was determined to be adequate for its intended function.

Concrete cores were taken from all Class I structure concrete in the affected areas, and Swiss hammer correlation calibration was made with standard cylinders from concrete being used on the project. A grid pattern was then established in each area to be tested (some 2 ft to 3 ft vertically by approximately 10 ft horizontally on centers in the biological shield area) with the calibrated Swiss hammer.

The Swiss hammer tests essentially measured the surface hardness of the concrete tested. Core tests furnished the depth and condition of the concrete in the structure. The correlation calibration was established to verify areas not cored. The Swiss hammer is also useful to check uniformity; to locate areas of unsatisfactory concrete in walls, beams, floors, and mass structures; and to serve as a substitute for test cylinders cured at the site for evaluating the compressive strength at early ages.

The Swiss hammer tests and core test were conducted in accordance with the requirements of ACI 301 Chapter 17, "Evaluation of Concrete Strength," paragraph 1704. The Swiss hammer used on affected areas was calibrated and correlated for each concrete sample tested (ACI 301, paragraph 1704(a)). Cores were taken and tested to ASTM C-42 (ACI 301, paragraph 1704(b)). More than three cores were taken in each affected area (ACI 301, paragraph 1704(c)). Approximately 10 impact hammer tests were taken within APPENDIX C C.4-10 REV. 27, APRIL 2019

PBAPS UFSAR a cored area. In accordance with ACI 301, paragraph 1704(d), an evaluation of stresses of affected areas was made and compared with core test results. The strength of all cores exceeded that required for the members with a safety factor equal to or greater than specified in the ACI Code.

Ultrasonic non-destructive testing was not employed in the affected areas as the cores gave actual strengths of concrete in place and Swiss hammer readings verified areas between cores.

Microfissures in lower than anticipated strength concrete considered satisfactory for compressive strength does not decrease the shear and bond resistance below a safe level.

This is demonstrated by the shear mode of break in the cores.

Thus the compressive strength of the core actually indicates the shear strength of the concrete.

Since bond transfer is mechanically done through reinforcing bar deformations, the compressive and shear strength of the concrete represents a measure of the bond also. Therefore, shear and bond strengths were lowered as the compressive strength was lowered but not below a safe level.

Recent research and testing in connection with microfissures in concrete has shown that microfissures exist before loading and stressing and are caused by normal settlement of aggregate, bleeding of the mixing water and shrinkage stresses induced by the drying process. Spreading of the cracks is retarded by interaction of surrounding particles, continuity of surrounding mass of concrete, roughness of aggregate, restraint of surrounding matrix, reinforcing steel, and pores and voids.

It has been established that for loads below about 30 percent of

' the increase in bond cracking is negligible. Above 30 percent fc and up to about 70 percent of fc' the amount of bond cracking increases causing the initial deviation from linearity of the stress-strain curve. Further cracking or increase in sizes is restricted until about 85 percent or 90 percent of ultimate is reached for the reasons outlined above.

Slow crack growth and propagation is associated with creep and shrinkage. Stress concentrations around the end of a microfissure invariably cause creep in the matrix until an equilibrium position is reached where stress concentrations are insufficient to propagate the microfissure. Therefore, under sustained loads APPENDIX C C.4-11 REV. 27, APRIL 2019

PBAPS UFSAR below ultimate, the microfissures stabilize. Besides, creep and shrinkage reduce the load carried by the mortar, further preventing the formation or propagation of a significant number of microfissures for loads below the ultimate.

Under biaxial or triaxial compressive stresses the strength is improved since normal stresses are restraining the propagation or formation of microfissures. Normal and shear stresses are interrelated by the classical Coulomb-Mohr theory, confirmed by tests.

As discussed above, the mode of failure for the concrete cores was in shear, not the hour-glass mode representative of compressive failures, thereby permitting the use of code factors for allowable stresses in structural concrete.

Actual stresses in the structures are well within those allowable for the core strengths tested.

Examination of microfissures indicated that there was no migration of gases; thus, it would not collect at rebar and have any more effect on the bond than on other characteristics. Therefore, bond is adequate.

References used are ACI journals of August 1964; January 1969; July 1969; and April 1971; and "Causes, Mechanism and Control of Cracking in Concrete," ACI publication SP-20.

APPENDIX C C.4-12 REV. 27, APRIL 2019

PBAPS UFSAR TABLE C.4.1 REACTOR BUILDING FLOOR SYSTEM Maximum Allowable Maximum Security Related Information Withheld under 1 Stress Stress Description/Criteria Method of Analysis Load Combination (ksi) (ksi)

• • This corner was designed Working stress design D + E Fc = 1.35 fc = 0.072 for two alternate method for steel and loadings: concrete Fv = 0.060 fv = 0.015 Uniformly distrib. Ft = 24.0 ft = 4.5 Live load of 1.0 ksf.

Fb = 24.0 fb = 17.6 Live load of RR car and 85T fuel cask Higher stresses due to either of the above govern the design Materials conform as D + E Fc = 2.55 fc = 0.076 follows:

Fv = 0.093 fv = 0.016 concrete fc = 3,000 psi at 28 days Ft = 32. ft = 4.9 concrete maximum strength per ACI 318-63 Fb = 32. fb = 18.4 Reinforcing ASTM Designation:

• A615 Grade 40 per ACI 318-63 Structural Steel ASTM Designation:

A 36-67 per AISC Manual and Speci-fication, 1963.

Note that this table provides stresses from the original analysis method to serve as an example of the typical load cases assessed and the results obtained.

• Reinforcing bars #8 and larger are A615 Grade 60. Other bars are A615 Grade 40 throughout the Class I structures.

• • Evaluation of this corner for other live load conditions is covered in next page.

APPENDIX C C.4-13 REV. 28, APRIL 2021

PBAPS UFSAR TABLE C.4.1 REACTOR BUILDING FLOOR SYSTEM Security Related Information Withheld un Maximum Allowable Maximum Allowable Max. Moment (k-ft)

Stress Moment or Description/Criteria Method of Analysis Load Combination (ksi) (k-ft) Max. Stress (ksi)

This corner was further evaluated for the following loading Working stress design method for steel and concrete.

D**

Fc = 1.35 Ft = 24.0 Mc = 525.4 M = 489.1 conditions:

Fv = 0.06 v = 0.056 Live load of cask transporter (70T) with Fb = 24.0 b = 14.0 a TN-68 cask (125T). The loading configuration from the TN-68 cask bounds the Holtec HI-TRAC VW loading Configuration.

Fe = 1.35 Same as above D+E Mc = 525.4 M = 515.0 Ft = 24.0 Fv = 0.06 v = 0.059 Fb = 24.0 b = 14.7 Same as above D+E Fe = 2.55 Ft = 32.0 Mc = 992.4 M = 551.7 Fv = 0.093 v = 0.063 Fb = 32.0 b = 15.8 Note that this table provides stresses from the original analysis method to serve as an example of the typical load cases assessed and the results obtained.

• Mc = Maximum allowable section bending moment capacity.

M = Maximum section bending moment

• • D = Includes both dead and live loads.

APPENDIX C C.4-14 REV. 28, APRIL 2021

PBAPS UFSAR TABLE C.4.2 REACTOR BUILDING CONCRETE WALLS Maximum Allowable Maximum Security Related Information Withheld under 10 Stress Stress Description/Criteria Method of Analysis Load Combination (ksi) (ksi)

Design load (D) Working stress D + E Ft = 24.0 ft = 18.2 includes all dead method and equipment loads Fc = 1.8 fc = 1.030 and soil pressure Fv = 0.070 fv = 0.053*

Materials conform as follows: Fv = 0.188 fv = 0.070**

Concrete Fc = 4,000 psi at 28 days D + E' Ft = 54. ft = 18.2 maximum strength per ACI 318-63 Fc = 3.40 fc = 1.13 Reinforcing ASTM Designation: Fv = 0.253 fv = 0.120 A615 Grade 60 per ACI 318-63 Maximum allowable stresses for D + E' Concrete Fc = .85 fc Reinforcing Ft = 0.9 fy

• Retaining wall shear

• • Seismic shear APPENDIX C C.4-15 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.4.3 REACTOR BUILDING STEEL SUPERSTRUCTURE Maximum Allowable Maximum Stress Stress Description/Criteria Method of Analysis Load Combination (ksi) (ksi) Location Material: Structural "STRESS" computer D + E Fa = -14.7 fa = -3.7 Corner columns Steel ASTM Designation: program using "stiff-A-36-67 per AISC ness" method for Fb = 24.0 fb = 1.0 Manual & Specifica- D & W' tion, 1963 D + W' Fa = -14.7 fa = -3.8 Corner columns Dynamic Analysis for Earthquake Fb = 32.0 fb = 1.8 Working Stress Design D + E Fa = 22.0 fa = 2.7 End column Method bracing D + W' Fa = 32.4 fa = 12.0 End column bracing D + W' Fa = 22.0 fa = 3.7 Center column Fb = 32.4 fb = 17.2 D + E Fa = -17.99 fa = -15.2 Truss top Chord D + W' Fa = -27.0 fa = -8.6 Truss bottom Chord D + E Fa = 22.0 fa = 13.2 Truss Diag.

D + W' Fa = -18.7 fa = -8.8 Truss Diag.

APPENDIX C C.4-16 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.4.4 REACTOR CONCRETE PEDESTAL Units 2 and 3 No Liners or Dowels Maximum Calculated Stresses (psi)

Maximum Allowable Stress @ El 119 ft 11 in @ El 130 ft 0 in Load Combination (psi) (Base)

D+T (Vert.) + E Rebar Tension .4Fy = 24,000 6,730 6,730 Compr. '

.45 f c 1,800 1,140 1,114 Concr. Shear** 1.1 f'c 70 41 41 Rebar Tension 24,000 7,440 7,440 D+T (Circ) + E Compr. 1,800 317 317 Concr. Shear** 70 70 70 Rebar Tension .9Fy = 54,000 7,345 7,345 D+T (Vert.) + E' Compr. ' =

.85 f c 3,400 1,161 1,114 Concr. Shear 4 f' c = 215 76 76 Rebar Tension 54,000 38,430 38,430 D+T (Vert.) + R Compr. 3,400 1,140 1,140 Concr. Shear 3.5 f' c = 188 58 58 Torsion* 10.2 f' c = 547* 326 326 Rebar Tension 54,000 40,400 40,400 D+T (Vert.) + E+R Compr. 3,400 1,585 1,555 Concr. Shear 215 96 96 Torsion* 547* 330 330 Rebar Tension 54,000 43,790 43,770 D+T (Vert.) + E'+R Compr. 3,400 1,602 1,555 Concr. Shear 215 131 131 Torsion* 547* 333 333 Unit 2 (two Liners)***

stresses not shown same as above.

Rebar 41,740 41,740 D + T + E' + R Bolts 17,065 Unit 3 (Dowels)***

Rebar 39,000 43,770 D + T + E' + R Dowels 39,000 -

APPENDIX C C.4-17 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.4.4 (Continued)

NOTES:

Description/Criteria The reactor vessel pedestal is a 25 ft 6 in high cylinder with 3 ft thick concrete walls and inside diameter of 20 ft 3 in. The pedestal projects from a spherical shaped base formed by the inside of the drywell. The shears and moments in the pedestal are transferred to the drywell through a welded steel shear ring and the bearing between the drywell and the concrete.

Temperature (T) = 70 F temperature gradient Design Earthquake (E) 0.05g, 2 percent damping.

Maximum Credible Earthquake (E') 0.12g, 5 percent damping Jet Force (R) = 1,000 K @ el 130 ft 0 in.

Materials: Concrete F c'= 4,000 psi at 28 days maximum strength per ACI 318-63, Reinforcing ASTM Designation: A615 Grade 60 per ACI 318-63.

Maximum allowable stresses for D + T + E' and D + T + R loads are: Concrete Fc = ' ; Reinforcing Ft =

0.85 f c 0.90 Fy. The customary increase in normal design stresses for other loading combinations when earthquake loads are considered is not used.

Methods of Analysis Working Stress Design Method.

For seismic loads response spectra are used.

Circumferential and vertical temperature analysis is in accordance with ACI 505-54.

• Torsional shear stress analysis based on "Tentative Recommendations for Design of Reinforced Concrete Members to Resist Torsion" by ACI Committee 438 in ACI Journal, January, 1969. Formula provides for interaction of flexural () and torsional () shears. Due to the openings in shell only 20 percent of full ring torsional constant was used in calculations of torsional shear stress.

Only 70 psi in shear are taken by concrete; rest by steel.

Tensile stresses in reinforcement are the sum of the stresses due to flexure and torsion.

• • Allowable stress is for unreinforced concrete. Radial ties are provided throughout the pedestal which would permit the use of higher allowable stresses (5 f' c )

= 316 .

• • • Taking base anchorage into account.

APPENDIX C C.4-18 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.4.5 DRYWELL SHIELDING CONCRETE Maximum Allowable Maximum Stress Description/Criteria Method of Analysis Load Combination* Stress (ksi) (ksi) Location Drywell shield acts as a structural Finite Element D + E Ft = 24.0 ft = 8.2 Circumferential El 145 ft wall carrying floors. Design load Stress Analysis reinforcing tension (D) consists of all dead loads, D + E Fc = 1.80 fc = 0.45 Vertical concrete equipment loads, and floor live compression load. D + E Fv = 0.07 fv = 0.0645 Concrete shear at El 180 ft Temperature stresses are for D + T + E' Ft = 54.0 ft = 23.5 Circumferential El 145 ft uncracked section. reinforcing tension Seismic loads (E and E') are D + T + E' Fc = 3.40 fc = 1.57 Vertical concrete according to the response spectra compression for the reactor building. D + T + E' Fv = 0.253 fv = 0.177 Concrete shear at El 180 ft Operational thermal load of 45 F "STRESS" computer program D + T + E'+ R Ft = 54.0 ft = 38.9 Circumferential El 141 ft (averaged) thermal gradient across using "stiffness" method reinforcing tension at main the wall is considered. Fc = 3.40 fc = 1.5 Vertical concrete steam compression penetrati Fv = 0.253 fv = 0.115 Concrete shear ons Materials conform as follows:

Concrete fc = 4,000 psi at 28 days maximum strength per ACI 318-63.

Reinforcing ASTM Designation: A615 Grade 60 per ACI 318-63.

Structural steel ASTM Designation A-36 per Specification and Manual of the AISC 6th Ed.

Maximum allowable stresses for D + T + E' load combination are:

Concrete Fc = 0.85fc Reinforcing Ft = 0.9fy

• Floor design load (D) includes dead and live loads plus 50-psi live loads (appropriate for operating conditions).

APPENDIX C C.4-19 REV. 25, APRIL 2015

PBAPS UFSAR C.5 COMPONENTS C.5.1 Intent and Scope C.5.1.1 Components Designed by Rational Stress Analysis These general design criteria are intended to apply to those ductile metallic structures or components which are normally designed using rational stress analysis techniques such as pressure vessels, reactor internal components, etc. The criteria may also be applied to those components or structures whose ultimate loading capability is determined by tests. These criteria are intended to supplement applicable industry design codes where necessary. Compliance with these criteria is intended to provide design safety margins which are appropriate to extremely reliable structural components when account is taken of rare event potentialities such as might be associated with an MCE or primary pressure boundary coolant pipe rupture, or a combination of events.

C.5.1.2 Components Designed Primarily by Empirical Methods There are many important seismic Class I components or equipment which are not normally designed or sized directly by stress analysis techniques. Simple stress anlayses are sometimes used to augment the design of these components, but the primary design work does not depend upon detailed stress analysis. These components are usually designed by tests and empirical experience.

Complete detailed stress analysis is currently not meaningful nor practical for these components. Examples of such components are valves, pumps, electrical equipment, and mechanisms. Field experience and testing are used to support the design. Where the structural or mechanical integrity of components is essential to safety, the components referred to in these criteria must be designed to accommodate the events of the MCE or design earthquake, or a design basis pipe rupture, or a combination where appropriate. The reliability requirements of such components cannot be quantitatively described in a general criterion because of the varied nature of each component and its specific function in the system.

C.5.1.3 Components Qualified for SQUG Methodology A specifically approved empirical method of equipment qualification is the use of seismic experience data utilized in accordance with the Seismic Qualification Utilities Group (SQUG) methodology to verify the seismic adequacy of existing, new, APPENDIX C C.5-1 REV. 27, APRIL 2019

PBAPS UFSAR modified and replacement items on a case-by-case basis. Such evaluations are performed in a controlled and systematic manner to ensure that the item of equipment is properly represented in the earthquake experience or generic testing classes and that applicable caveats are met. In particular, each new or replacement item must be evaluated for any design changes that could reduce the seismic capacity of the equipment from that reflected in the experience data base, and all such evaluations must be documented in accordance with established procedures. SQUG methodology is applied in accordance with the SQUG Generic Implementation Procedure (Reference 16) and implementation of the SQUG methodology is controlled and documented in accordance with approved procedures. The use of the SQUG methodology is limited to the scope of equipment covered by equipment classes described in the SQUG Generic Implementation Procedure (GIP). The methodology is not used to verify the seismic adequacy of equipment not included within the scope of the equipment classes described in the GIP, and may not be used when the NRC commitment has been made to qualify specific equipment to IEEE 344-75 (invoked by Reg.

Guide 1.100).

C.5.2 Loading Conditions and Allowable Limits The loading conditions established herein are expressed in generic terms and are related in a probabilistic manner to the loads which are to be investigated for safety considerations. Related probabilistic definitions are used to determine an appropriate minimum safety factor which is used to establish structural design allowable limits and functional design allowable limits. Certain of the limits described in these criteria, i.e., deformation limit and fatigue limit, are included for completeness, but do not necessarily require application to all components. Where it is clear to the designer that fatigue or excess deformation are not of concern for a particular structure or component, a formal analysis with respect to that limit is not required.

The design loading conditions which were used for components of this plant (except the reactor vessel and reactor vessel internals) are presented in Tables C.5.7 and C.5.8.

C.5.2.1 Loading Conditions The loading conditions may be divided into four categories:

normal, upset, emergency, and faulted conditions. No categorization for loading conditions was made for the safety/relief valves, safety valves, main steam line isolation valves, recirculation system valves and pumps, or other components APPENDIX C C.5-2 REV. 27, APRIL 2019

PBAPS UFSAR in the reactor coolant pressure boundary since the applicable codes did not require such categorization. The categories listed above are generically described as follows.

C.5.2.1.1 Normal Conditions Any condition in the course of operation of the station under planned and anticipated conditions, in the absence of upset, emergency, or faulted conditions.

• C.5.2.1.2 Upset Conditions Any deviations from normal conditions anticipated to occur often enough that design should include a capability to withstand those conditions. The upset conditions include abnormal operational transients caused by a fault in a system component requiring its isolation from the system, transients due to loss of load or power, and any system upset not resulting in a forced outage. The upset conditions may include the effect of the design earthquake.

For recirculation systems, the design earthquake, not the MCE, is the upset condition.

C.5.2.1.3 Emergency Conditions Any deviations from normal conditions which require shutdown for correction of the conditions or repair of damage in the system.

The conditions have low probability of occurrence, but are included to provide assurance that no gross loss of structural integrity will result as a concomitant effect of specific damage developed in the system.

C.5.2.1.4 Faulted Conditions Those combinations of conditions associated with extremely low probability postulated events whose consequences are such that the integrity and operability of the nuclear system may be impaired to the extent where considerations of public health and safety are involved. Such considerations require compliance with safety criteria. The faulted condition includes the effects of the MCE.

• See Note A APPENDIX C C.5-3 REV. 27, APRIL 2019

PBAPS UFSAR

• C.5.2.2 Allowable Limits In addition to the generic definition of loading conditions in the preceding paragraphs, the meaning of these terms is expanded in the quantitative probabilistic language. The purpose of this expansion is to clarify the classification of any hypothesized accident or sequence of loading events so that the appropriate limits or safety margins are applied. Knowledge of the event probability is necessary to establish meaningful and adequate safety factors for design. The quantitative event classifications are as follows:

Loading Conditions Probabilities Upset (likely) 1.0 > P40 10-1 Emergency (low probability) 10-1 > P40 10-3 Faulted (extremely low probability) 10-3 > P40 10-6 where P40 = 40-yr event encounter probability These probabilities have been assigned to establish the appropriate structural design limits for the loading conditions in paragraph C.5.2.1. A summary of these limits is shown in the tables listed below:

Deformation Limit Table C.5.1 Primary Stress Limit Table C.5.2 Buckling Stability Limit Table C.5.3 Fatigue Limit Table C.5.4 There are many places where, through the exercise of designer judgment, it is unnecessary to actually carry out a formal analysis for each of these limits. A simple example consists of the case where two pieces of pipe of different wall thicknesses are joined at a butt weld. If they are both subjected to the same loading, only the thinner piece would require a formal analysis to demonstrate that the primary stress limit has been satisfied.

• See Note A APPENDIX C C.5-4 REV. 27, APRIL 2019

PBAPS UFSAR The other SM is defined as the minimum safety factor on load or deflection and is related to the event probability by the following equation:

SFmin = 9 3-log10 P40 where:

10-1 > P40 10-5 For event probabilities smaller than 10-5 or greater than 10-1, the following apply:

Event Probability Min. Safety Factor 10-5 > P40 > 10-6 1.125 1.0 > P40 > 10-1 2.25 These expressions show the probabilistic significance of the classical safety factor concept as applied to reactor safety. The SFmin values corresponding to the event probabilities are summarized in Table C.5.5.

The loadings which occur as a result of the conditions listed are factored into the design of the components in accordance with the requirements of the applicable design code, or to the requirements of these criteria. Where permitted by the applicable code and by these criteria, the SFmin may be progressively lowered to a minimum acceptable level on the basis that there is a lesser need for design margin for loading conditions which have a diminishing probability of occurrence.

• NOTE A:

In Table C.5.5 of the Peach Bottom 2 and 3 FSAR, the 40-year event encounter probability for the Maximum Credible Earthquake (MCE) plus normal loads is given to be 10-3. Using the currently accepted probability categories given in Table 1, MCE plus peak operating loads is a faulted condition. In the replacement recirculation piping analysis and Main Steam piping analysis APPENDIX C C.5-5 REV. 27, APRIL 2019

PBAPS UFSAR inside containment, MCE plus peak operating loads was analyzed as a faulted condition.

In the original analysis, MCE plus peak operating loads was analyzed as a faulted condition and MCE plus normal operating loads was analyzed as an emergency condition.

The following tables do not apply to recirculation and Main Steam piping inside containment:

Table C.5.2 Table C.5.4 Table C.5.5 The Main Steam piping inside containment meet Table C.5.7 requirement.

(NOTE A CONTINUED)

Table 1 Current Probability Categories* and Acceptance Criteria for Components other than Containment Structures Normal Conditions P1 ~ 1.0 ----> P40 = 40 Upset Conditions 1 > P1 10-2 40 > P40 0.4 Emergency Conditions 10-2 > P1 10-4 0.4 > P40 4 x 10-3 Faulted Conditions 10-4 > P1 10-6 4 x 10-3 > P40 4 x 10-5 P1 = 1-year event encounter probability P40 = 40-year event encounter probability APPENDIX C C.5-6 REV. 27, APRIL 2019

PBAPS UFSAR Safe Shutdown Earthquake (SSE) is characterized as having an encounter probability of <10-4 per reactor year. This is equal to an encounter probability of <4 x 10-3 in 40 reactor years.

• For BWR 4, 5, & 6 reactors utilizing Mark I, II, and III containments.

APPENDIX C C.5-7 REV. 27, APRIL 2019

PBAPS UFSAR C.5.3 Method of Analysis and Implementation of Criteria C.5.3.1 Reactor Vessel The reactor pressure vessels are designed, fabricated, inspected, and tested in accordance with Section III of the 1965 Edition of the ASME Boiler and Pressure Vessel Code as described in Appendix K.

The ASME code does not require categorization of loading conditions.

Stress analysis requirements and load combinations for the reactor vessel have been evaluated for the cyclic conditions expected throughout the 60-year life, with the conclusion that ASME code limits are satisfied. The monitoring locations for the cycles used in the evaluation of these vessels is presented in Table 4.2.4. The results of the original stress analysis are presented in Table C.5.6. References 4, 5, 6 and 35 document the re-evaluation of reactor vessel fatigue in accordance with the CUF of the locations presented in Table 4.2.4 prior to implementation of Extended Power Uprate. Reconciliation of reactor vessel stress and fatigue evaluation performed for EPU is presented in References 19, 20, and 22, with the conclusion that ASME code limits for the affected components continue to be satisfied.

The vessel design report contains the results of the detailed design stress analyses performed for the reactor vessel to meet the code requirements. Selected components considered to possibly have higher than code design primary stresses as a result of rare events or a combination of rare events have been analyzed in accordance with the requirements of the loading criteria in this appendix.

Results of the most critical of those original analyses are included in Table C.5.6. The conclusion is that the limits in the criteria have been met. Results of reactor vessel stress reconciliation for EPU are presented in Reference 19.

Closure stresses and usage factors have been re-evaluated in M-1-A-411 based on a reduced number of tensioning and detensioning passes for RPV assembly and disassembly.

C.5.3.1.1 Vessel Fatigue Analysis An analysis of the reactor vessel shows that all components are adequate for cyclic operation by the rules of Section III of the ASME Code. The critical components of the vessel are evaluated on a fatigue basis, calculating cumulative usage factors (ratios of required cycles to allowed cycles-to-failure) for all operating cycle conditions. The cumulative usage factors for the critical components of the vessel are below the code allowable of 1.0.

APPENDIX C C.5-8 REV. 27, APRIL 2019

PBAPS UFSAR References 19 and 20 document the fatigue analysis of the affected reactor pressure vessel components for EPU operating conditions, including re-analysis of the feedwater nozzle. The analysis considers effects of environmentally assisted fatigue due to reactor coolant environment for a 60-year plant life (Reference 18). For the feedwater nozzle, cumulative usage factor is based on both system and rapid cycling effects on the inner radius. The cumulative usage factors for the critical components are below the code allowable limit.

C.5.3.1.2 Vessel Seismic Analysis A seismic analysis was performed for a coupled system consisting of reactor building, drywell, reactor vessel, and internals. The analysis is discussed in paragraph C.5.3.2.

C.5.3.2 Reactor Vessel Internals Although not mandatory, the design of the reactor vessel internals is in accordance with the intent of Section III of the ASME Boiler and Pressure Vessel Code. Most of the material used for fabrication is solution heat-treated, unstabilized type 304 austenitic stainless conforming to ASTM specifications. Allowable stresses for the internals materials under normal operating conditions are taken directly from Section III. The methods of analysis used as a guide were the design procedures of Section III. For rare events or a combination of rare events, the internals have been analyzed in accordance with the requirements of the loading criteria in this appendix, and results of the most critical of those original analyses are included in Table C.5.6.

The conclusion is that the limits in the criteria have been met.

Analysis of affected reactor vessel internals for EPU conditions is documented in References 21 and 22, with the conclusion that ASME Section Ill code limits for allowable stress and fatigue usage are satisfied.

C.5.3.2.1 Internals Deformation Analysis Control Rod System If there were excessive deformation of the CRDS, made up of the CRD, CRD housing, control rod, control rod guide tube and fuel channels, and the core structural elements which support them (top guide, core support, shroud, and shroud support), it could possibly impede control rod insertion. The maximum loading condition that would tend to deform these long, slender components APPENDIX C C.5-9 REV. 27, APRIL 2019

PBAPS UFSAR is the MCE. Analyses of the internal components which have the highest calculated stresses are included in a subsequent paragraph. The highest calculated stresses occur where the MCE and loads resulting from the design basis accident line break are considered to occur simultaneously. Even in these cases, the general stress levels are relatively low. No significant deformation is associated with these calculated stresses; therefore, rod insertion would not be impeded after an assumed simultaneous MCE and line break accident. Reconciliation analysis of the control rod system performed for EPU continues to support the conclusion that rod insertion would not be negatively affected. (Reference 21)

Core Support The core support sustains the pressure drop across the fuel. The pressure drop is the only load which causes significant deflection of the core support. Excessive core support deflection could lift the control rod guide tubes off their seats on the CRD housings and thereby increase core bypass leakage. This upward deflection would have to be 1/2 in to begin to lift the guide tubes. The maximum deflections under normal operation conditions and pipe rupture differential pressures for the core support are calculated to be very small as compared to 1/2 in. The guide tubes will, therefore, not be lifted off, although even if they were, this would not be of concern because bypass leakage at this time is not important. Reconciliation analysis performed for EPU continues to support this conclusion for the core support (Reference 19).

C.5.3.2.2 Internals Fatigue Analysis Fatigue analysis was performed using as a guide the ASME Boiler and Pressure Vessel Code,Section III. The method of analysis used to determine the cumulative fatigue usage is described in General Electric Topical Report APED-5460, "Design and Performance of GE-BWR Jet Pumps," September, 1968. The most significant fatigue loading occurs in the jet pump - shroud - shroud support area of the internals. The analysis was performed for a plant where the configuration (leg type shroud support) was almost identical to the Peach Bottom plant. Therefore, the calculated fatigue usage is expected to be a reasonable approximation for this plant.

Loading Combinations and Transients Considered

1. Normal startup and shutdown APPENDIX C C.5-10 REV. 27, APRIL 2019

PBAPS UFSAR

2. Design and MCE's

3. Ten minute blowdown from a stuck relief valve

4. HPCI operation

5. LPCI operation (design basis accident)

6. Improper start of a recirculation loop Conclusion The cumulative fatigue usage factor for Peach Bottom is evaluated to be less than 1 (Ref. 5). Based on the reconciliation analysis performed for EPU, the cumulative usage factor for the affected internals is less than the ASME Code limit of 1.0 (Reference 21).

Remarks The location of maximum fatigue usage is at the bottom side of the baffle plate at the point where the baffle plate attaches to the shroud in the vicinity of the minimum ligament.

C.5.3.2.3 Internals Seismic Analysis The seismic loads on the reactor vessel and internals are based on a dynamic analysis of the coupled model consisting of reactor building, reactor vessel, and internals. The natural frequencies and mode shapes for the system were determined. The relative displacement, acceleration, and load response of the reactor vessel and internals were then determined using the time-history method of analysis. The dynamic response was determined for each mode of interest and added algebraically for each instant of time.

Resulting response time-histories were then examined, and the maximum value of displacements, accelerations, shears, and moments were used for design calculations. These results were combined with the results of other loads for the various loading conditions. The combined results for the critical components from the original analysis are presented in Table C.5.6. Seismic analysis of the internals is not affected by EPU implementation.

Increased weight of the replacement steam dryer is reconciled as documented in PEAM-EPU-130 (Reference 23). For additional details see Appendix K, Exhibit V.

C.5.3.3 Piping APPENDIX C C.5-11 REV. 27, APRIL 2019

PBAPS UFSAR C.5.3.3.1 Piping Flexibility Analysis The piping has been analyzed for the effects of dead loads, external loads, and thermal loads. In addition, piping attached to the torus has also been analyzed for the effects of Safety Relief Valve (SRV) discharge loads and loss of coolant accident (LOCA) loads which consists of pool swell, chugging, and condensation oscillation. Stresses calculated are combined bending and torsional stresses in accordance with ANSI B31.1, "Power Piping," and intensification factors were applied in accordance with ANSI B31.1. (See Section A.1.1) Several pressure/temperature cycles were evaluated, and the cycle representing the worst for thermal expansion stresses was selected for the design case. All critical points were evaluated to the stress limits of the design code and, in addition, events with a very low probability of occurrence were analyzed and stresses at all critical points compared with the limits defined in this loading criteria. Additionally, the effects of Mark I cyclic mechanical loads on torus attached piping was addressed in Report MPR-751, "Augmented Class 2/3 Fatigue Evaluation Method and Results for Typical Torus Attached and SRV Piping Systems" dated November 1982. Results showed that fatigue usage factors are low, typically below 0.3 and all below 0.5 compared to a code limit of 1.0 for a plant lifetime.

The recirculation and RHR piping systems were designed to ASME-III Class 1 (see Section A.1.1), and Torus Attached Piping was evaluated in accordance with the codes listed in Section A.1.1.

In the course of design progress it has been determined that the weld reinforcement limit criteria given in Appendix A are not applicable to the systems requiring in-service inspection in accordance with ASME Boiler and Pressure Vessel Code,Section XI.

In order to meet the requirements of ultrasonic examination for in-service inspection of welds within the primary coolant pressure boundary (Group I), the weld reinforcement heights are equal to or less than required in ANSI B31.1. In addition, the main steam lines from downstream of the outer isolation valves to the main turbine stop valves and the feedwater lines from the pump discharge to the reactor coolant pressure boundary meet the requirements of ANSI B31.1.

Class II and the project design requirements as stated in Appendix A. The load combination, allowable stresses, and identification of points of highest stress are summarized in Table C.5.7.

APPENDIX C C.5-12 REV. 27, APRIL 2019

PBAPS UFSAR C.5.3.3.2 Piping Seismic Analysis Piping, 2 inch and smaller, is analyzed by one of three methods:

the span chart method, the simplified static method, or the computer method. The simplified static method and the computer method are described in Appendix A, paragraph A.3.1.4. Piping, 2-1/2 in and larger, is dynamically analyzed by the response spectrum method described in paragraph A.3.1.4 of Appendix A. For each of the piping systems, a mathematical model consisting of lumped masses at discrete joints connected together by weightless elastic elements was constructed. Valves were also considered as lumped masses in the pipe, and valve operators as lumped masses acting through the operator center of gravity. Where practical, a support is located on the pipe at or near each valve. Stiffness matrix and mass matrix were generated and natural periods of vibration and corresponding mode shapes were determined. Input to the dynamic analyses were the acceleration response spectra for the applicable floor elevations. For the design earthquake, 0.5 percent damping is used. For the maximum credible earthquake, 0.5 percent damping is used for Unit 2 and 1.0 percent damping is used for Unit 3. The increased flexibility of the curved segments of the piping systems was also considered. Except for the 1997 re-analysis of the Recirculation system piping, and the Residual Heat Removal and Reactor Water Clean-up piping inside primary containment for Peach Bottom NCR 97-02267, the results of earthquakes acting in the X and Y (vertical) directions simultaneously, and Z and Y directions simultaneously were computed separately. The maximum responses of each mode are calculated and combined by the root-mean-square method to give the maximum quantities resulting from all modes. The response thus obtained was combined with the results produced by other loading conditions to compute the resultant stresses.

Some Torus attached piping were seismically analyzed based on the response spectra at the piping center of mass although other pipe portions are located and supported from higher elevations. An analysis of piping stress calculation S/11187/D-68, which represents the worst case, or enveloping condition of all configurations per ECR 01-00077, was performed by Bechtel Corporation. The analysis was performed using NRC approved code case N-411 with an envelope response spectra of all applicable elevations. The re-analysis results supported acceptability of the original center of mass methodology. Based on results of the analysis, no re-analysis of other calculations are required using actual floor elevations. Therefore, those stress calculations which utilized center of mass spectra are concluded to be acceptable and will continue to utilize this approach in future APPENDIX C C.5-13 REV. 27, APRIL 2019

PBAPS UFSAR re-analyses. Those calculations are for piping stress analyses of systems 10 (RHR), 12 (RWCU), 13 (RCIC), 14 (CS), and 23 (HPCI) and their numbers are S/11187/D-050, S/11187/D-051, S/11187/D-053, S/11187/D-059, S/11187/D-060, S/11187-015/D-064, S/11187/D-067, S/11187/D-068, S/11187/D-070, S/11187/D-072, S/11187/D-073, S/11187/D-077, S/11187/D-080, S/11187/D-096, 10-39, 10-41, 23-04, and 23-9.

The 1997 re-analysis of the Recirculation system piping, and the Residual Heat Removal and Reactor Water Clean-up piping inside the primary containment for Peach Bottom NCR 97-02267 combined the peak collinear contributions due to the three spatial components of seismic excitation by the square root-sum of the squares (SRSS) method as required by the application of ASME Code Case N-411-1.

In this method, separate analyses are conducted corresponding to the three spatial components (two horizontal, one vertical) of seismic excitation resulting in an analysis of a three (3) dimensional earthquake.

As indicated in paragraph C.1.2 and Appendix A, each main steam line up to and including the main steam line isolation valve external to the primary containment is seismic Class I. The main steam line anchor is in the seismic Class I portion of the main steam line and, therefore, does not separate the seismic Class I part of the main steam line from the seismic Class II part.

Additional analysis of the main steam lines from outer main steam isolation valves up to but not including the turbine stop valves indicates that because they are restrained from the steam tunnel to the turbine, including restraint for fast valve closure, the lines will meet the stress requirements of seismic Class I piping as defined in Appendix A. Additional restraints are provided for the seismic Class II portions of the main steam lines to protect the adjacent seismic Class I piping in the pipe tunnel. The design methods and design stress criteria are similar to those provided for Monticello Unit 1, AEC Docket No. 50-263.

The main steam line anchor at the penetration adapter was designed to resist dead load, thermal loads, and design earthquake loads within normal AISC code allowable stresses. The anchor was also designed to withstand dead load, thermal loads, MCE, and the design accident loads within allowable stresses as discussed in paragraph C.3.2.

All seismic restraints and snubbers were located after a dynamic analysis determined their necessity. After the supports and restraints were installed, they were field checked to ensure compliance with the assumptions in the analysis.

APPENDIX C C.5-14 REV. 27, APRIL 2019

PBAPS UFSAR C.5.3.3.3 Piping Mark I Load Analysis Torus attached piping greater than 4-inch NPS has been dynamically analyzed using displacement time-history analysis by the modal superposition method for safety relief valve and pool swell loads.

Displacement time-histories were developed and applied at the torus nozzle for each degree of freedom (three displacements and three rotations). Mass point spacing was selected based upon a maximum significant frequency of 50 Hz. In addition, for piping close to the torus nozzle, additional mass points were selected to obtain significant responses. The damping factors used are shown in Table C.5.9.

The harmonic analysis method was used to analyze the piping systems which were subjected to condensation oscillation and chugging loads. Fifty (50) individual analyses were run for chugging for frequencies from 1 to 50 Hz. The results of each of the 50 responses were summed absolutely. Thirty individual analyses were performed for condensation oscillation for frequencies from 1 to 30 Hz. The four largest results were summed absolutely and added to the remaining twenty-six which were combined by square-root-sum-of-the-squares.

For torus attached piping 4-inch NPS and less, static analysis was performed using a dynamic load factor of 2.0. Six displacements (three translations and three rotations) were prescribed at each torus nozzle. The analysis of branch piping was similarly analyzed except the rotational displacements were neglected as they have a translation displacement equivalent of less than 1/16 inch.

The evaluation of Mark I hydrodynamic loads is in accordance with the provisions of Section A.1.1.

C.5.3.4 Equipment The extent of stress analyses performed on equipment is dependent upon the type of equipment and the type of fabrication.

Fabricated shapes are generally made from plate or rolled shapes with uniform thickness and shapes with regular geometric configurations. Cast shapes are generally made with non-uniform material thickness in complicated shapes that are not regular geometric configurations. Manufacturers have traditionally designed cast shapes conservatively since they do not lend themselves to rational analysis. Usually a design is developed based on extensive test and experience. The equipment was APPENDIX C C.5-15 REV. 27, APRIL 2019

PBAPS UFSAR analyzed to determine equipment adequacy for earthquake loading.

The equivalent static coefficients for equipment were obtained from applicable floor response spectra corresponding to the support elevations of the equipment. In lieu of determining the natural frequency of the equipment, the peak value of the applicable floor response spectrum was used in calculating the earthquake induced loads. Alternately, the natural frequency of the equipment was determined and corresponding input acceleration was obtained from the appropriate floor response spectra. The criteria, method of analysis, and summary of critical stresses for various equipment are included in Table C.5.8.

For existing, new, modified or replacement equipment installed at Peach Bottom, the Seismic Qualification Utilities Group (SQUG) methodology may be used in lieu of the methodology described above to verify the seismic adequacy of the equipment on a case-by-case basis. SQUG methodology is applied in accordance with the SQUG Generic Implementation Procedure (Reference 16) and implementation of the SQUG methodology shall be controlled and documented in accordance with approved procedures.

C.5.3.5 Cable Trays Cable trays, battery racks, instrument racks, and control consoles which are by definition seismic Class I (paragraph C.1.1),

considering the safety functions required, are supported or restrained to withstand, without loss of safety functions, the effects of the MCE (horizontal ground acceleration of 0.12g).

The design of the cable tray support systems is the product of extensive investigation of hanger systems. Design adequacy is verified by dynamic analysis. Battery racks are designed for static coefficients (0.24g), and the adequacy of these coefficients confirmed by dynamic analysis. Instrument racks and control consoles are dynamically analyzed and restrained for natural frequencies equal to or greater than 20 Hz.

C.5.3.6 Reactor Coolant System Supports Recirculation piping stresses were calculated in accordance with the ASME B&PV Code,Section III, Article NB-3600, 1980 Edition, up to and including Winter 1981 addenda. The load combinations and allowables are shown in Table C.5.7. The following transients are considered in the stress analyses of the recirculation piping:

Transient Category Cycles APPENDIX C C.5-16 REV. 27, APRIL 2019

PBAPS UFSAR Startup/Shutdown Normal 216 Turbine roll and increase Normal 216 to power Loss of feedwater heater Upset 10 Partial feedwater heater Upset 70 bypass Scrams Upset 180 Loss of feedwater pumps, Upset 10 isolation valves closed Reactor overpressure Emergency 1 with delayed scram Single SRV blowdown Upset 8 Automatic Blowdown Emergency 1 Hydrotest Test 226 OBE - design Normal/upset 50 All component supports for the recirculation piping, main steam piping (to the first anchor outside the drywell), and the remainder of the reactor coolant systems were designed to the codes in effect at the time the purchase order was placed. The design, materials, and fabrication of parts were in accordance with Power Piping Code, ANSI B31.1, and the Standard MSS-SP-58, as applicable. The suspension systems were designed in accordance with the criteria presented below. This table applies to both variable and constant support hangers and to seismic restraints.

Design conditions, load combinations, and calculated stress for the recirculation system pipe whip restraints are presented in Table C.5.6.

The design of new component supports and parts for recirculation pump snubbers and new hanger clamps, in conjunction with recirculation pipe replacement, was in accordance with ASME B&PV Code,Section III, Subsection NF, 1980 Edition, up to and including Winter 1981 addenda. The materials and fabrication of parts for recirculation pump snubbers and new hanger clamps were in accordance with ASME B&PV Code,Section III, Subsection NF, 1980 Edition, up to and including Winter 1980 addenda for Peach Bottom 2, and Section III, Subsection NF, 1980 Edition, up to and including Winter 1981 addenda for Peach Bottom 3.

Ambient Conditions Temperature 70F (prior to initial startup) 135F normal/150F maximum (during operation and shutdown)

APPENDIX C C.5-17 REV. 27, APRIL 2019

PBAPS UFSAR Relative Humidity 40% (during operation) 95% (during shutdown)

Radiation 100 Rads/hr (3.5 x 107 R/40 yr)

Primary Membrane Load Combinations Stress Limits Weight + Thermal S(1)

Expansion, Design Earthquake Weight + Thermal 0.9 Sy(3)

DESIGN OF PEACH BOTTOM 2 AND 3 RECIRCULATION PUMP SNUBBERS New piping supports installed in conjunction with the recirculation pipe replacement program are designed in accordance with Subsection NF of the ASME B&PV Code,Section III. Supports are either designed by load rating, per Subsection NF-3260, or to the stress limits for linear supports, per Subsection NF-3231. To avoid buckling in the component supports, Appendix F of the ASME B&PV Code requires that the allowable loads be limited to two-thirds of the critical buckling loads. The critical buckling loads that are more severe than normal, upset, and emergency loads, are determined by the supplier, using the methods discussed in Appendix F of the ASME B&PV Code. In general, the load combinations used for the design of component supports correspond to those used to design the supported pipe. Design transient cyclic data are not applicable to piping supports, since no fatigue evaluation is necessary to meet the ASME B&PV Code requirements.

Stresses in the snubber component supports under normal, upset, emergency, and faulted loads are calculated. These calculated stresses are then compared against the allowable stresses of the material, as given in the ASME B&PV Code,Section III, to make sure that they are below the allowable limits.

APPENDIX C C.5-18 REV. 27, APRIL 2019

PBAPS UFSAR C.5.3.7 References

1. NEDC-32183P, "Power Rerate Safety Analysis Report for Peach Bottom 2 & 3", May 1993

2. GE NE-123-E239-1292, REV 1, "Peach Bottom Atomic Power Plant Units 2 and 3 105% Power Rerate Evaluation of Main Steam and Recirculation Piping," Class II, Dec 1993.

3. NEDC-32230P, "Peach Bottom Power Rerate Project Engineering Report" Class III, March 1994.

4. GE-NE-523-61-0493, "Fatigue Evaluation of the Peach Bottom II and III Reactor Vessels," May 1993.

5. GE Letter, GENE B13-01805-73, Oct. 18, 1996.

6. GE Letter, WFW 9607, Nov. 13, 1996.

7. Deleted

8. 23A4065, Peach Bottom Unit 2, GE Design Specification for Recirculation System Piping Section XI Replacement (PECO Document Number M-1-U-499)

9. 23A4608, Peach Bottom Unit 3, GE Design Specification for Recirculation System Piping Section XI Replacement (PECO Document Number M-1-U-502)

10. 23A4086, Peach Bottom Unit 2, GE Certified Design Report for Recirculation System Piping and Equipment Loads, Loop A (PECO Document Number M-1-U-504)

11. 23A4645, Peach Bottom Unit 2, GE Certified Design Report for Recirculation System Piping and Equipment Loads, Loop B (PECO Document Number M-1-U-505)

12. 23A8026, Peach Bottom Unit 3, GE Certified Design Report for Recirculation System Piping and Equipment Loads, Loop A (PECO Document Number M-1-U-506)

13. 23A8027, Peach Bottom Unit 3, GE Certified Design Report for Recirculation System Piping and Equipment Loads, Loop B (PECO Document Number M-1-U-507)

APPENDIX C C.5-19 REV. 27, APRIL 2019

PBAPS UFSAR

14. NEDC-33064P Safety Analysis Report for Peach Bottom Atomic Power Station Units 2 & 3 Thermal Power Optimization.

15. NEDC-32938P Licensing Topical Report: Generic Guidelines and Evaluations for General Electric Boiling Water Reactor Thermal Power Optimization.

16. Seismic Qualification Utilities Group (SQUG) Generic Implementation Procedure (GIP) Revision 3A.

17. G-080-VC-313, "Project Task Report, Peach Bottom Atomic Power Station Unit 2 and 3, SIL 636 Evaluation," GE-NE-0000-011-4483, Rev. 0, Class III, March 2003.

18. "Project Task Report, Peach Bottom Atomic Power Station Unit 2 and 3, Extended Power Uprate task T0400: Containment System Response," GE-0000-0130-9920-R1, Rev. 1, August 2012.

19. G-080-VC-411, T0302, RPV- Stress and Fatigue Evaluation, PBAPS Units 2 and 3.

20. PEAM-EPU-128, T0302, RPV - Environmentally Assisted Fatigue Analysis (SIA), PBAPS Units 2 and 3

21. PEAM-EPU-9, T0303, RPV Internals Mechanical Evaluation for EPU, PBAPS Units 2 and 3

22. G-080-VC-423; NEDC-33566P, Safety Analysis Report for Exelon Peach Bottom Atomic Power Station, Units 2 and 3, Constant Pressure Power Uprate, Revision 0, September 2012

23. PEAM-EPU-130, RSD Disposition on GEH Task Evaluations, PBAPS Units 2 and 3

24. PEAM-EPU-29, T0400 Containment System Response, PBAPS Units 2 and 3

25. Calculation PS-1042, "Development of Primary Structure N-S and E-W Seismic Models"

26. Calculation PS-1043, "Development of Primary Structure N-S and E-W Taft In-Structure Spectra for MSL Piping Analysis"

27. Calculation 1-31, "Piping Stress Analysis for Main Steam Line A Inside Containment and HPCI Line" APPENDIX C C.5-20 REV. 27, APRIL 2019

PBAPS UFSAR

28. Calculation 1-32, "Piping Stress Analysis for Main Steam Line B Inside Containment and HPCI Line"

29. Calculation 1-33, Piping Stress Analysis for Unit 2 Main Steam Line C Inside Containment and HPCI Line"

30. Calculation 1-34, "Piping Stress Analysis for Main Steam Line D Inside Containment and HPCI Line"

31. Calculation 1-35, "Piping Stress Analysis for Unit 3 Main Steam Line C Inside Containment and HPCI Line"

32. Calculation 1-36, "Piping Stress Analysis for Unit 3 Main Steam Line A Inside Containment and HPCI Line"

33. Calculation 1-37, "Piping Stress Analysis for Unit 3 Main Steam Line B Inside Containment and HPCI Line"

34. Calculation 1-38, "Piping Stress Analysis for Unit 3 Main Steam Line D Inside Containment and HPCI Line"

35. Calculation PM-1164, Peach Bottom Environmentally Assisted Fatigue Screening.

36. NEI 17-02, Tornado Missile Risk Evaluator (TMRE) Industry Guidance Document, Rev. 1A, July 2018.

37. Calculation PS-1238 Rev. 0, Evaluation of Peach Bottom Atomic Power Station's Tornado Missile Protection Design for compliance with the licensing requirements.

38. 168-J-VC-1, PBAPS TMRE Model Development and Quantification, Rev. 0.

APPENDIX C C.5-21 REV. 27, APRIL 2019

PBAPS UFSAR TABLE C.5.1 DEFORMATION LIMIT Either One of (Not Both) General Limit Permissible Deformation , DP 09

a.

Analyzed deformation SFmin DL cau sin g loss of function, Permissible Deformation , DP 10

b.

Experimental deformation SFmin causing loss of function, DE where:

DP = permissible deformation under stated conditions of normal, upset, emergency, or fault DL = analyzed deformation which would cause a system loss of function*

DE = experimentally determined deformation which would cause a system loss of function*

• "Loss of Function" can only be defined quite generally until attention is focused on the component of interest. In cases of interest, where deformation limits can affect the function of equipment and components, they will be specifically delineated.

From a practical viewpoint, it is convenient to interchange, with the loss of function condition, some deformation condition at which function is assured if the required safety margins from the functioning condition can be achieved. Therefore, it is often unnecessary to determine the actual loss of function condition because this interchange procedure produces conservative and safe designs. Examples where deformation limits apply are: CRD alignment and clearances for proper insertion, core support APPENDIX C C.5-22 REV. 21, APRIL 2007

PBAPS UFSAR deformation causing fuel disarrangement, or excess leakage of any component.

APPENDIX C C.5-23 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.2 PRIMARY STRESS LIMIT Any One of (No More than One Required) General Limit Elastic Evaluated Primary Stresses, PE 2.25

a. SFmin Permissible Primary Stresses, PN

• Permissible Load, LP 1.5

b. Largest Lower Bound Limit Load, CL

• SFmin Elastic Evaluated Primary Stress, PE 0.75

c.

Conventional Ultimate Strength at SFmin Temperature, US

• Elastic - Plastic Evaluated Nominal Primary Stress, EP

• 0.9

d. Conventional Ultimate Strength at SFmin Temperature, US

• Permissible Load, LP 0.9

e. Plastic Instability Load, PL

• SFmin Permissible Load, LP 0.9

f. Ultimate Load from Fracture SFmin Analysis, UF

• Permissible Load, LP 1.0

g. Ultimate Load or Loss of Function SFmin Load from Test, LE

• where:

PE = Primary stresses evaluated on an elastic basis. The effective membrane stresses are to be averaged through the load carrying section of interest. The simplest average bending, shear, or torsion stress distribution, which will support the external loading, will be added to the membrane stresses at the section of interest.

PM = Permissible primary stress levels under normal or upset conditions under applicable industry code.

LP = Permissible load under stated conditions of emergency or fault.

APPENDIX C C.5-24 REV. 21, APRIL 2007

PBAPS UFSAR

• See NOTES, Table C.5.3.

APPENDIX C C.5-25 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.3 BUCKLING STABILITY LIMIT Any One of (No More than One Required) General Limit

a. Permissible Load, LP 2.25 Code normal event permissible load, PN SFmin

b. Permissible Load, LP 0.674 Stability Analysis Load, SL SFmin

c. Permissible Load, LP 1.0 Ultimate Buckling Collapse Load from SFmin Test, SE where:

NOTES LP = Permissible load under stated conditions of normal, upset, emergency, or fault.

PN = Applicable code normal event permissible load.

SL = Stability analysis load. The ideal buckling analysis is often sensitive to otherwise minor deviations from ideal geometry and boundary conditions. These effects shall be accounted for in the analysis of the buckling stability loads. Examples of this are ovality in externally pressurized shells or eccentricity of column members.

SE = Ultimate buckling collapse load as determined from experiment. In using this method, account shall be taken of the dimensional tolerances which may exist between the actual part and the tested part. The guide to be used in each of these areas is that the experimentally determined load shall be adjusted to account for material property and dimension variations, each of which has no greater probability than 0.1 of being exceeded in the actual part.

CL = Lower bound limit load with yield point equal to 1.5 Sm where Sm is the tabulated value of allowable stress at temperature as contained in ASME Section III or its equivalent. The lower bound limit load is here defined as that produced from the analysis of an ideally plastic (non-strain APPENDIX C C.5-26 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.3 (Continued) hardening) material where deformations increase with no further increase in applied load. The lower bound load is one in which the material everywhere satisfies equilibrium and nowhere exceeds the defined material yield strength using either a shear theory or a strain energy of distortion theory to relate multiaxial yielding to the uniaxial case.

US = Conventional ultimate strength at temperature or loading which would cause a system malfunction, whichever is more limiting.

EP = Elastic plastic evaluated nominal primary stress. Strain hardening of the material may be used for the actual monotonic stress strain curve at the temperature of loading or any approximation to the actual stress strain curve, which everywhere has a lower stress for the same strain as the actual monotonic curve, may be used.

Either the shear or strain energy of distortion flow rule may be used.

PL = Plastic instability load. The plastic instability load is defined here as the load at which any load bearing section begins to diminish its cross-sectional area at a faster rate than the strain hardening can accommodate the loss in area. This type analysis requires a true stress-true strain curve or a close approximation based on monotonic loading at the temperature of loading.

UF = Ultimate load from fracture analyses. For components which involve sharp discontinuities (local theoretical stress concentration > 3), the use of a fracture mechanics analysis where applicable, utilizing measurements of plane strain fracture toughness, may be applied to compute fracture loads. Correction for finite plastic zones and thickness effects as well as gross yielding may be necessary. The methods of linear elastic stress analysis may be used in the fracture analysis where its use is clearly conservative or supported by experimental evidence. Examples where fracture mechanics may be applied are for fillet welds or end of fatigue life crack propagation.

LE = Ultimate load or loss of function load as determined from experiment. In using this method, APPENDIX C C.5-27 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.3 (Continued) account shall be taken of the dimensional tolerances which may exist between the actual part and the tested part or parts as well as differences which may exist in the ultimate tensile strength of the actual part and the tested parts. The guide to be used in each of these areas is that the experimentally determined load shall use adjusted values to account for material properties and dimension variations, each of which has no greater probability than 0.1 of being exceeded in the actual part.

APPENDIX C C.5-28 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.4 FATIGUE LIMIT Summation of mean a. Fatigue cycle .05(2) fatigue(1) damage usage from usage including analysis emergency or fault events with design and b. Fatigue cycle 0.33 operation loads following usage from test Miner's Hypotheses...either one (not both)

(1) Fatigue failure is defined here as a 25 percent area reduction for a load carrying member which is required to function, whichever is more limiting. In the fatigue evaluation, the methods of linear elastic stress analysis may be used when the 3S range limit of ASME Section III has been met. If 3S is not met, account will be taken of (a) increases in local strain concentration, (b) strain ratcheting, and (c) re-distribution of strain due to elastic-plastic effects. The January, 1969 draft of the USAS B31.7 Piping Code may be used where applicable or detailed elastic-plastic methods may be used. With elastic-plastic methods, strain hardening may be used not to exceed in stress for the same strain, the steady-state cyclic strain hardening measured in a smooth low cycle fatigue specimen at the average temperature of interest.

(2)

It is acceptable to use the ASME Section III Design Fatigue curves in conjunction with a cumulative usage factor of 1.0 (using Miner's Hypothesis) in lieu of using the mean fatigue APPENDIX C C.5-29 REV. 21, APRIL 2007

PBAPS UFSAR data curves with a limit on fatigue usage of 0.05, since the two methods are approximately equivalent.

APPENDIX C C.5-30 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.5 MINIMUM SAFETY FACTORS Loading Conditions Loads P40 SFmin Upset N and AD 10-1 2.25 or N and U 10-1 2.25 Emergency N and R 10-3 1.5 N and Am 10-3 1.5 Other combi- <10-1 to 10-3 <2.25 to 1.5 nations in this proba-bility range Fault N and Am 1.5 x 10-6 1.125 and R Other combi- <10-3to 10-6 <1.5 to 1.125 nations in this proba-bility range KEY:

N = Normal loads U = Upset loads (result in maximum system pressure)

AD = Design earthquake Am = MCE R = Loads resulting from jet forces and pressure and temperature transients associated with rupture of a single pipe within the primary containment.

This load is considered as indicated in the tables.

NOTE: The minimum safety factor decreases as the event probability diminishes, and if the event is too APPENDIX C C.5-31 REV. 21, APRIL 2007

PBAPS UFSAR improbable (incredible:P40 < 10-6), then the safety factor is appropriate or required.

APPENDIX C C.5-32 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.6 REACTOR VESSEL AND REACTOR VESSEL INTERNALS Criteria Loading Primary Stress Type Allowable Stress Calculated Stress*

Stabilizer Bracket and Adjacent Shell Primary Stress Limit - ASME Normal and upset condi- Membrane and bending 40,000 psi 29,700 psi Boiler and Pressure Vessel Code, tion load Sect. III, defines primary mem- 1. Design earthquake brane plus primary bending 2. Design pressure stress intensity limit for SA Emergency condition load Membrane and bending 60,000 psi 32,200 psi 302 - Gr. B 1. MCE For normal and upset condition 2. Design pressure Stress limit = 1.5x26,700 = Faulted condition loads 40,000 psi 1. MCE Membrane and bending 80,000 psi 35,100 psi For emergency condition 2. Jet reaction forces Stress limit = 1.5x40,000 = 3. Design pressure 60,000 psi For faulted condition Stress limit = 2.0x40,000 =

80,000 psi Vessel Support Skirt Primary Stress Limit - ASME Normal and upset condi- General membrane 26,700 psi 6,600 psi Boiler and Pressure Vessel Code, tion loads Sect. III, defines stress limit 1. Dead weight for SA 302, Gr. B 2. Design earthquake Emergency condition loads For normal and upset condition 1. Dead weight General membrane 40,000 psi 8,100 psi Sm = 26,700 psi 2. MCE For emergency condition Faulted condition loads Slimit = 1.5 Sm = 1.5x26,700 = 1. Dead weight General membrane 53,400 psi 11,900 psi 40,000 psi 2. MCE For faulted condition 3. Jet reaction forces Slimit = 2.0 Sm = 2.0x26,700 =

53,000 psi

• These results represent original plant design. Analysis of power rerate conditions is documented in References 1 and 2. Analysis for Thermal Power Optimization is presented in Reference 14 and 15 APPENDIX C C.5-33 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.6 (Continued)

Criteria Loading Primary Stress Type Allowable Stress Calculated Stress Shroud Leg Support Primary Stress Limit - ASME Normal and upset condi- Tensile 23,300 psi 12,600 psi Boiler and Pressure Vessel Code, tion loads Sect. III, defines allowable 1. Design earthquake primary membrane stress SB-168 2. Pressure drop across material shroud (normal)

1. Tensile Loads 3. Subtract deadweight For normal and upset con- Emergency condition loads Tensile 35,000 psi 22,900 psi dition 1. MCE Sm = 23,300 psi 2. Pressure drop across shroud (normal)

For emergency condition 3. Subtract dead weight Slimit = 1.5 Sm = Faulted condition loads

1. MCE Tensile 46,600 psi 28,200 psi 1.5x23,300 = 35,000 psi 2. Pressure drop across For faulted condition shroud during faulted Slimit = 2.0 Sm = condition

3. Subtract dead weight 2.0x23,300 = 46,600 psi

2. Compressive Loads For normal and upset con- Normal and upset condi- Compressive 14,000 psi 12,500 psi ditions tion loads SA = 0.4 Sy = 0.4x35,000 = 1. Design earthquake 14,000 psi 2. Zero pressure drop a For emergency condition across shroud Slimit = 0.6 Sy = 0.6x 3. Dead weight 35,000 = 21,000 psi Emergency condition loads Compressive 21,000 psi 17,000 psi

1. MCE For faulted condition 2. Subtract operating Slimit = 0.8 Sy = 0.8x pressure drop across 35,000 = 28,000 psi shroud

3. Dead weight Faulted condition loads Compressive 28,000 psi 26,800 psi

1. MCE

2. Zero pressure drop across shroud

3. Dead weight APPENDIX C C.5-34 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.6 (Continued)

Criteria Loading Primary Stress Type Allowable Stress Calculated Stress Top Guide-Longest Beam Primary Stress Limit - The Normal and upset condi- General membrane plus 24,000 psi 21,600 psi allowable primary membrane tion loads bending stress plus bending stress 1. Design earthquake is based on ASME Boiler and 2. Weight of structure Pressure Vessel Code, Sect. 3. Weight of temporary III, for type 304 stainless control curtain steel plate.

For normal and upset condi- Emergency condition loads General membrane plus 36,000 psi 30,700 psi tion Stress Intensity 1. MCE bending SA = 1.5 Sm = 1.5x16,000 psi = 2. Weight of structure 24,000 psi 3. Weight of temporary For emergency condition control curtains Slimit = 1.5 SA = 1.5x24,000 psi =

36,000 psi Faulted condition loads General memberane plus 48,000 psi 30,700 psi (same as emergency bending For faulted condition condition)

Slimit = 2SA = 2x24,000 =

48,000 psi Top Guide Beam End Connections Primary Stress Limit - ASME Normal and upset condi- Pure shear 9,600 psi 9,100 psi Boiler and Pressure Vessel Code, tion loads Sec. III, defines material 1. Design earthquake stress limit for type 304 2. Weight of structure stainless steel. 3. Weight of temporary control curtains For normal and upset condition Stress Intensity SA = 0.6 Sm =

0.6x16,000 psi = 9,600 psi. Emergency condition loads Pure shear 14,400 psi 14,400 psi

1. MCE For emergency condition 2. Weight of structure Slimit = 1.5 SA = 1.5x9,600 psi 3. Weight of structure control curtains

= 14,000 psi For faulted condition Slimit = 2SA = 2x9,600 psi = Faulted condition loads Pure shear 19,200 psi 14,400 psi 19,200 psi (same as emergency condition)

APPENDIX C C.5-35 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.6 (Continued)

Criteria Loading Primary Stress Type Allowable Stress Calculated Stress Core Support Primary Stress Limit - The Normal and upset condi- General membrane plus 24,000 psi 15,250 psi allowable primary membrane tion loads bending stress plus bending stress 1. Normal operation is based on ASME Boiler and pressure drop Pressure Vessel Code, Sect. 2. Design earthquake III for type 304 stainless steel plate Emergency condition loads General membrane plus 36,000 psi 22,500 psi For allowable stresses see 1. Normal operation bending top guide, longest beam, pressure drop above. 2. MCE Faulted condition loads General membrane plus 48,000 psi 26,500 psi

1. Pressure drop after bending recirculation line rupture

2. MCE Core Support Aligners Primary Stress Limit - ASME Normal and upset condi- Pure shear 9,600 psi 0 Boiler and Pressure Vessel tion load Code, Sect. III, defines 1. Design earthquake material stress limit for type 304 stainless steel.

For allowable shear stresses, Emergency condition load Pure shear 14,400 psi 12,000 psi see top guide beam end 1. MCE connections above.

Faulted condition load Pure shear 19,200 psi 12,000 psi

1. MCE APPENDIX C C.5-36 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.6 (Continued)

Moment Limit Accounting for Criteria Loading Primary Stress Type Pressure Loads Maximum Moment Fuel Channels Primary Stress Limit - Allow- Normal and upset condi- Membrane and bending 28,230 in-lb 6,927 in-lb able stress Sm for Zircaloy tion load determined according to methods 1. Design earthquake recommended by ASME Boiler and 2. Normal pressure load Pressure Vessel Code, Sect. III.

Allowable moment determined by calculating limit moment using Emergency condition load Membrane and bending 42,350 in-lb 16,625 in-lb Table C.3.2, equation (b), then 1. MCE applying SFmin for applicable 2. Normal pressure load loading conditions.

(Sm = 9,270 psi; 1.5 Sm = Faulted condition load Membrane and bending 56,500 in-lb 16,625 in-lb 13,900 psi) 1. MCE Emergency limit load = 1.5 x 2. Loss of cooling acci-Normal limit load calculated dent pressure using 1.5 Sm = yield.

APPENDIX C C.5-37 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.6 (Continued)

Criteria Loading Location Allowable Stress Calculated Stress RPV Stabilizer Primary Stress Limit Upset condition Rod 136,000 psi ft = 62,400 psi*

AISC specification for the 1. Spring preload construction, fabrication, 2. Design earthquake Bracket 22,000 psi fb = 15,500 psi and erection of structural steel for buildings 14,000 psi fv = 4,600 psi For normal and upset conditions Emergency condition Bracket 33,000 psi fb = 20,000 psi AISC allowable stresses, but 1. Spring preload without the usual increase for 2. MCE 21,000 psi fv = 6,000 psi earthquake loads For emergency conditions Faulted Condition Bracket 36,000 psi fb = 21,500 psi 1.5 x AISC allowable 1. Spring preload stresses 2. MCE 21,500 psi fv = 6,400 psi For faulted conditions 3. Jet reaction load Material yield strength

• The ratio maximum stress/stress limit is highest for upset loading conditions RPV Support (Ring Girder)

Primary Stress Limit Normal and upset condi- Top Flange 27,000 psi fb = 15,000 psi AISC specification for the tion design fabrication and 1. Dead loads erection of structural steel 2. Design earthquake for buildings 3. Loads due to scram Bottom Flange 27,000 psi fb = 14,100 psi Vessel to girder bolts 60,000 psi ft = 35,200 psi For normal and upset con 22,500 psi fv = 4,450 psi ditions AISC allowable stresses, but without the usual increase for earth-quake loads For faulted conditions Faulted condition Top Flange 45,000 psi fb = 39,500 psi 1.67 x AISC allowable stresses 1. Dead loads for structural steel members 2. MCE Bottom Flange 45,000 psi fb = 37,500 psi Yield strength for high strength 3. Jet reaction Load bolts (vessel to ring girder) Vessel to girder 125,000 psi ft =115,000 psi bolts 75,000 psi fv = 11,800 psi APPENDIX C C.5-38 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.6 (Continued)

Criteria Loading Location Allowable Stress Calculated Stress CRD Housing Support Primary Stress Limit Faulted condition loads Beams (Top cord 33,000 psi fa = 15,300 psi AISC specification for the design, fabrication and 1. Dead weight 33,000 psi fb = 12,700 psi erection of structural steel for buildings 2. Impact force from Beams (bottom cord) 33,000 psi fa = 11,800 psi failure of a CRD housing 33,000 psi fb = 7,600 psi For normal and upset Grid structure* 41,500 psi fa = 40,000 psi condition 27,500 psi fb = 11,100 psi Fa = 0.60 Fy (tension)

Fb = 0.60 Fy (bending)

Fv = 0.40 Fy (shear For faulted conditions Fa limit = 1.5 Fa (tension)

Fb limit = 1.5 Fb (bending)

Fv limit = 1.5 Fv (shear)

Fy = Material yield strength

• (Dead weights and earthquake loads are very small as compared to jet force.)

Recirculating Pipe and Peach Bottom Unit 3 Pump Restraints Information under this heading deleted.

APPENDIX C C.5-39 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.6 (Continued)

Criteria Loading Location Allowable Stress Calculated Stress THIS PAGE DELETED APPENDIX C C.5-40 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.6 (Continued)

Criteria Loading Location Allowable Stress Calculated Stress Control Rod Guide Tube Primary Stress Limit Faulted condition loads The maximum bending 47,400 psi 7,535 psi The allowable primary membrane 1. Dead weight stress under faulted stress plus bending stress is 2. Pressure drop across loading conditions based on the ASME Boiler and guide tube due to occurs at the center Pressure Vessel Code, Sect. III, failure of recircu- of the guide tube.

for Type 304 stainless steel lation line tubing 3. MCE For normal and upset conditions SA = 1.5 Sm = 1.5x15,800 =

23,700 psi For faulted condition Slimit = 2.0 SA = 2.0x23,700 =

47,400 psi In-core Housing Primary Stress Limit - The Emergency condition loads Maximum membrane 23,700 psi 15,290 psi allowable primary membrane 1. Design pressure stress intensity stress is based on ASME 2. MCE occurs at the outer Boiler and Pressure Vessel surface of the vessel Code, Sect. III, for Class A penetration.

vessels for Type 304 stainless steel For normal and upset conditions Sm = 15,800 psi at 575 F For emergency condition (N+Am)

Slimit = 1.5 Sm = 1.5x15,800 =

23,700 psi APPENDIX C C.5-41 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.6 (Continued)

Criteria Loading Location Allowable Stress Calculated Stress Fuel Storage Racks Stresses due to normal, upset, Emergency condition At column to base 11,000 psi 9,620 psi(1) or emergency loading shall not "A" loads welds cause the racks to fail so as 1. Dead loads to result in a critical fuel 2. Full fuel load in At base hold down 20,000 psi 16,600 psi(2) array. rack lug (casting)

3. MCE Primary Stress Limit - Paper numbers 3341 and 3342, pro-ceedings of the ASCE, Journal Emergency condition of the Structural Division, "B" loads(3)

Dec. 1962 (task committee on light-weight alloys) (Aluminum)

Emergency Conditions Stress limit = yield strength at 0.2% offset.

(1)Load testing showed that the structure would not yield when subjected to simulated emergency condition "A" loads.

Strain gages mounted on the welds showed that calculated stresses are conservative.

(2)Calculated stresses compare very well with test results.

(3)Emergency Condition "B" Loading In addition to the loading conditions given above, the racks were tested and analyzed to determine their capability to safely withstand the accidental, uncontrolled drop of the fuel grapple from its full retracted position into the weakest portion of the rack.

Method of Analysis The displacement of the vertical columns at the ends of the racks was determined by considering the effect of the grapple kinetic energy on the upper structure. The energy absorbed shearing the rack longitudinal structural member welds was determined.

The effect of the remaining energy on the vertical columns was analyzed. Equivalent static load tests were made on the structure to assure that the criteria were met.

APPENDIX C C.5-42 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.6 (Continued)

Results of Analysis All criteria were met.

Analysis showed that the grapple would shear the welds in the area where the impact occurred. The longitudinal structural member bends but does not fail in shear. Grapple penetration into the rack is not sufficient to cause the vertical columns to deflect the fuel into a critical array. Static load testing showed that forces in excess of those resulting from a grapple drop are required to cause the columns to deflect to the extent that the criteria is violated.

APPENDIX C C.5-43 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.7 LOADING COMBINATIONS AND CRITERIA FOR PEACH BOTTOM 2 AND 3 Main Steam Piping Operating Condition / Service Design Basis Load Case Combinations Code Equation Stress Allowable Level 1.0 Sh Sustained Peak Pressure + Deadweightsteam EQN. 11 (at design temperature) 1.0 Sh Sustained Peak Pressure + Deadweightwater EQN. 11 (at design temperature)

Occasional Peak Pressure + Deadweightsteam + Design Earthquake EQN. 12 1.2 Sh (Upset) (DE) Level B (at design temperature)

Occasional EQN. 12 1.2 Sh Peak Pressure + Deadweightsteam + TSV Transient (Upset) Level B (at design temperature)

Occasional EQN. 12 1.2 Sh Peak Pressure + Deadweightsteam + RV Transient (Upset) Level B (at design temperature)

Occasional Peak Pressure + Deadweightsteam + SRSS [Design EQN. 12 1.8 Sh (Emergency) Basis Earthquake (DE) and TSV Transient] Level C (at design temperature)

Occasional Peak Pressure + Deadweightsteam + SRSS [Design EQN. 12 1.8 Sh (Emergency) Basis Earthquake (DE) and RV Transient] Level C (at design temperature)

Peak Pressure + Deadweightsteam + SRSS [Maximum Occasional EQN. 12 2.4 Sh Credible Earthquake (MCE) and TSV Transient and (Faulted) Level D (at design temperature)

RV Transient]

Thermal Stress Range + DE Seismic Anchor SA Thermal Expansion EQN. 13 Movements (SAM) (SA = f (1.25 Sc + 0.25 Sh))

Peak Pressure + Deadweightsteam + Thermal Stress Sustained + Thermal Expansion EQN. 14 (Sh + SA)

Range + SAM For the occasional load the pressure associated with the event is used to obtain stresses.

The material allowable stress Sh and Sc are based on a factor of safety of 3.5 based on Evaluation 01137961-04 (EC EVAL 388022).

The stresses for the Main Steam line inside containment for lines A, B, C, and D are documented in Reference 31 through 34 and all stresses are within the allowables per applicable codes.

APPENDIX C C.5-44 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.7 (Continued)

STEAM TO HPCI TURBINE Criteria Loading Allowable Stress Calculated Stress Secondary Stress The sum of the thermal 1. Thermal expansion stress and 22,500 psi 6,617 psi earthquake equipment 2. Earthquake displacement stress equipment should conform to al- displ.

lowable stress as stress given in B31.1.

Primary Stress The sum of the longi- 1. Internal tudinal stresses due pressure to internal pressure, 18,000 psi 12,028 psi dead weight, and iner- 2. Dead tia effects of a de- weight sign earthquake should be less than 1.2 times 3. Design the hot allowable earthquake stress.

Primary Stress The sum of the longi- 1. Internal tudinal stresses due pressure to internal pressure, 26,000 psi 17,561 psi dead weight, and iner- 2. Dead tia effects of an MCE weight should be less than the hot yield stress. 3. MCE NOTE: 1. The calculated stress is the sum of the various maximum stresses, which do not necessarily all occur at the same point in the piping system.

Therefore, comparing this calculated stress with the allowable stress is a conservative procedure.

2. The 105% rerate condition raised the thermal stress by 2% and the pressure stress by 4%.

Overall stress is still well within the allowable stress.

3. Maximum thermal expansion stress for Unit 3 HPCI piping inside the torus compartment from Modification P00634 is 17,315 psi.

APPENDIX C C.5-45 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.7 (Continued)

CORE SPRAY PUMP SUCTION DELETED APPENDIX C C.5-46 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.7 (Continued)

ASME CODE CLASS 1 PEACH BOTTOM 2 RECIRCULATION PIPING - HIGHEST STRESS LOCATION Identification of Locations of Calculated Ratio Highest Stress Limiting Stress(1) or Allowable Actual/ Points - NODG Acceptance Criteria Stress Type Usage Factor Limits Allowable Loading Point Numbers ASME B&PV Code Section III, NB-3600 Design condition:

1. Pressure Loop A

2. Weight Lug discharge (096)

Eq. 9 1.5 Sm Primary 24760 psi 25875 psi 0.96 3. DE Service levels A & B (normal & upset) condition:

Eq. 12 3.0 Sm Secondary 36010 psi 51750 psi 0.70 3. Thermal expansion Loop A Riser Tee (062)

Service levels A & B Primary 45171 psi 51750 psi 0.87 1. Pressure Loop A (normal & upset) plus secon- 2. Weight Disch. Riser condition: dary (except 3. DE Reducer (062) thermal Eq. 13 3.0 Sm expansion)

Service levels A & B 1. Pressure Loop A (normal and upset) 2. Weight Header extruded condition: 3. Thermal expansion Outlet (085)

4. DE Cumulative usage factor 0.109 1.0 0.109 Loop A Disch.Riser Reducer (062)

Loop B Riser Tee (085)

Service level B (upset) Primary 25219 psi 29388 psi 0.86 1. Pressure Loop A condition: 2. Weight Lug discharge (096)

3. DE Eq. 9 1.8 Sm

& .15 Sy Service level C (emergency) condition: None APPENDIX C C.5-47 REV. 25, APRIL 2015

PBAPS UFSAR APPENDIX C C.5-48 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.7 (Continued)

ASME CODE CLASS 1 PEACH BOTTOM 2 RECIRCULATION PIPING - HIGHEST STRESS LOCATION Identification of Locations of Calculated Ratio Highest Stress Limiting Stress(1) or Allowable Actual/ Points - NODG Acceptance Criteria Stress Type Usage Factor Limits Allowable Loading Point Numbers Service level D (faulted) condition:

1. Pressure Loop A

2. Weight Lug Discharge (096)

Eq. 9 < 3.0 Sm Primary 26970 psi 39184 psi 0.69 3. MCE

& 2.0 Sy APPENDIX C C.5-49 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.7 (Continued)

ASME CODE CLASS 1 PEACH BOTTOM 3 RECIRCULATION PIPING - HIGHEST STRESS LOCATION Identification of Locations of Calculated Ratio Highest Stress Limiting Stress(1) or Allowable Actual/ Points - NODG Acceptance Criteria Stress Type Usage Factor Limits Allowable Loading Point Numbers ASME B&PV Code Section III, NB-3600 Design condition:

1. Pressure Loop A

2. Weight Lug discharge (181)

Eq. 9 1.5 Sm Primary 23962 psi 25875 psi 0.93 3. DE Service levels A & B (normal & upset) condition:

Eq. 12 3.0 Sm Secondary 24073 psi 51750 psi 0.47 3. Thermal expansion Loop A RWCU/Recirc.

Tee (500)

Service levels A & B Primary 41779 psi 51750 psi 0.79 1. Pressure Loop A (normal & upset) plus secon- 2. Weight Recirc. Pump condition: dary (except 3. DE Suction Elbow (47) thermal Eq. 13 3.0 Sm expansion)

Service levels A & B 1. Pressure Loop A (normal and upset) 2. Weight RHR Tee (602) condition: 3. Thermal expansion

4. DE Cumulative usage factor 0.009 1.0 0.009 Loop A RHR/Recirc. Tee (500)

Service level B (upset) 1. Pressure Loop A condition: 2. Weight Lug discharge (181)

3. DE Eq. 9 1.8 Sm Primary 24400 psi 29388 psi 0.83

& .15 Sy Service level C (emergency) condition: None APPENDIX C C.5-50 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.7 (Continued)

ASME CODE CLASS 1 PEACH BOTTOM 3 RECIRCULATION PIPING - HIGHEST STRESS LOCATION Identification of Locations of Calculated Ratio Highest Stress Limiting Stress(1) or Allowable Actual/ Points - NODG Acceptance Criteria Stress Type Usage Factor Limits Allowable Loading Point Numbers Service level D (faulted) condition:

1. Pressure Loop A

2. Weight Suction Elbow (045F)

Eq. 9 < 3.0 Sm Primary 29261 psi 39184 psi 0.75 3. MCE

& 2.0 Sy APPENDIX C C.5-51 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.7 (Continued)

LOADING COMBINATIONS AND CRITERIA FOR PEACH BOTTOM 2 AND 3 RECIRCULATION ASME B&PV CODE CLASS 1 PIPING AND COMPONENTS The loading combinations given below were considered and the calculated stresses are reported for the governing load.

ASEM B&PV Code Event(1) Service Limit PD + W + DE Design N A N + OBE + SOT B None C Pp + W + MCE D (1) Key to load definitions:

PD = Design Pressure W = Weight N = Normal load consisting of pressure, dead weight, and thermal loads.

DE = Design earthquake SOT = Systems operating transients MCE = Maximum credible earthquake PP = Peak system operating pressure APPENDIX C C.5-52 REV. 25, APRIL 2015

PBAPS UFSAR TABLE C.5.8 EQUIPMENT MAIN STEAM ISOLATION VALVES Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required

1. Body Minimum Wall Thickness Minimum wall thickness in the Body wall thickness cylindrical portions of the valve Loads: shall be calculated using the following formula: t = 1.875 in t = 1.83 at 23" diam Design pressure & temperature Pd Primary Membrane Stress t = 15. + C 2S 12

.P Limit:

S = 7,000 lb/in2 per ASA B16.5 where:

S = allowable stress of 7,000 psi P = primary service pressure, 655 psi d = inside diameter of valve at section being considered, in C = corrosion allowance of 0.12 in

2. Cover Minimum Thickness Valve cover thick-ness and stress CP . Whg 178 1 2

Loads: t = d + + C1 S Sd3 Design pressure & temperature where:

Design bolting load Gasket load t = minimum thickness, in t = 5.469 in t = 4.888 in d = diameter or short span, in Primary Stress Limit: C = attachment factor Sallow = 17,800 lb/in2 S = allowable stress, psi Allowable working stress W = total, bolt load, lb per ASME Section VIII hG = gasket moment arm, in C1 = corrosion allowance, in APPENDIX C C.5-53 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

MAIN STEAM ISOLATION VALVES (Continued)

Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required

3. Cover Flange Bolt Area: Total, bolting loads and stresses shall Flange Bolt Area &

be calculated in accordance with "Rules Stress Loads: for Bolted Flange Connections" - ASME S = 30,900 lb/in2 Boiler & Pressure Vessel Code, Section at 575 F Ab = 55.29 in2 Design Pressure & temperature VIII, Appendix II, except that the stem Gasket load operational load and seismic loads shall Sb = 15,400 lb/in2 Stem operational load be included in the total load carried Seismic load - maximum by bolts. The horizontal and vertical credible earthquake seismic forces shall be applied at the mass center of the valve operator assum-ing that the valve body is rigid and Bolting Stress Limit: anchored.

Allowable working stress per ASME Nuclear Pump &

Valve Coe, Class I

4. Body Flange Thickness & Flange thickness and stress shall be Body Flange Thick-Stress calculated in accordance with "Rules ness and Stress for Bolted Flange Connections" - ASME Loads: Boiler & Pressure Vessel Code, Section for t = 4 in VIII, Appendix II, except that the S = 26,700 lb/in2 SH = 25,520 lb/in2 Design pressure & temperature stem operational load and seismic S = 26,700 lb/in2 SR = 14,120 lb/in2 Gasket load loads shall be included in the total S = 26,700 lb/in2 ST = 4,900 lb/in2 Stem operational load load carried by the flange. The Seismic load - maximum horizontal and vertical seismic forces credible earthquake shall be applied at the mass center of the valve operator assuming that the Flange Stress Limits: valve body is rigid and anchored.

SH, SR, ST, 1, 5, Sm per ASME Nuclear Pump

& Valve Code, Class I.

5. Valve Disc Thickness Sr = St = 3 (3 + ) PR2 Valve Disc Thick-8 t2 ness and Stress Loads:

for t = 3.563 in Design pressure & temperature where: Sr = St = 16,830 lb/in2 Sr = radial stress, psi S = 17,800 lb/in2 St = tangential stress APPENDIX C C.5-54 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

MAIN STEAM ISOLATION VALVES (Continued)

Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required Primary bending stress = Poisson's ratio limit: P = design pressure, psi R = radius of disc, in Allowable working stress t = thickness of disc, in per ASME Section VIII

6. Valve Operator Supports The valve assembly shall be analyzed Operator Support assuming that the valve body is an Stress & Deflection Loads: anchored, rigid mass and that the specified vertical and horizontal S = 18,000 lb/in2 Combined bending &

Design pressure & temper- seismic forces are applied at the tensile stress ature mass center of the operator assembly S = 5,400 lb/in2 Stem operational load simultaneously with operating pressure Equipment dead weight plus dead weight plus operational Seismic load - maximum loads. Using these loads, stresses Deflection at Operator credible earthquake and deflections shall be determined for the operator support components. S = 0.032 in Support Rod Stress Limit:

Allowable working stress per ASME Section VIII.

APPENDIX C C.5-55 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

MAIN STEAM SAFETY VALVES Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required

1. Inlet Nozzle Wall Thickness PR Loads: t = + C SE 06

.P 1.1 x Design Press. at 600 F where:

Primary Membrane Stress t = min. required thickness, in t = 0.784 in t = 0.183 in Limit: S = allowable stress, lb/in2 P = 1.1 x design press., lb/in2 Allowable stress intensity R = internal radius, in as defined by ASME Standard E = joint efficiency Code for Pumps & Valves for C = corrosion allowable, in Nuclear Power

2. Valve Disc Thickness W PA1 Ss = =

A A Loads:

1.1 x Design Press. at 600 F where:

Diagonal Shear Stress Limit: W = shear load, lb Ss = 20,190 lb/in2 S = 13,617 lb/in2 A = shear area, in2 0.6 x allowable stress in- P = 1.1 x design press., lb/in2 tensity as defined by ASME A1 = disc area, in2 Standard Code for Pumps &

Valves for Nuclear Power. and:

A = S (R + R1)

S = slope of frustrum of shear cone, in R = radius at base of cone, in R1 = radius at top of cone, in APPENDIX C C.5-56 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

MAIN STEAM SAFETY VALVES Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required

3. Inlet Flange Bolt Area Total bolting loads and stresses shall be calculated in accordance with pro-Loads: cedures of Para. 1-704.5.1 Flanged Joints, of B31.7 Nuclear Piping Code. Sb = 27,700 lb/in2 Sb = 17,297 lb/in2 Design pressure & temper-ature Gasket load Operational load Maximum credible earthquake Bolting Stress Limit:

Allowable stress intensity, Sm , as defined by ASME Stand-ard Code for Pumps & Valves for Nuclear Power

4. Inlet Flange Thickness Flange thickness and stresses shall SH = 27,300 lb/in2 SH = 21,339 lb/in2 be calculated in accordance with SR = 27,300 lb/in2 SR = 10,798 lb/in2 Loads: procedures of Para. 1-704.5.1 ST = 27,300 lb/in2 ST = 4,581 lb/in2 Flanged Joints, of B31.7 Design pressure & temper- Nuclear Piping Code.

ature Gasket load Operational load Seismic load - maximum credible earthquake Flange Stress Limits:

SH , SR , ST 1.5 Sm per ASME Nuclear Pump & Valve Code APPENDIX C C.5-57 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

MAIN STEAM SAFETY VALVES (Continued)

Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required

5. Valve Spring - Torsional Set Point Set Point 8PD 4C 1 0615 Stress Smax = = S = 82,500 lb/in2 S = 65,693 lb/in2 d3 4C 4 C Loads:

Torsional Stress Limit:

W1 = Set point load, lbs where:

W2 = Spring load at maxi- S = 112,500 lb/in2 S = 112,500 lb/in2 mum lift, lb Smax = torsional stress, lb/in2 P = W1 or W2 = spring load, lb 0.67 x torsional elastic D = mean diameter of coil, in limit when subjected to a d = diameter of wire, in load of W1 C = D = correction factor d

0.90 x torsional elastic limit when subjected to a load of W2.

6. Yoke Rod Area A = F Loads: 2sm Spring load at maximum where:

lift A = required area per rod, in2 A = 2.67 in2 A = 0.852 in2 Primary Stress Limit: F = total spring load, lb Sm = allowable stress, lb/in2 Allowable stress inten-sity, Sm, as defined by ASME Standard Code for Pumps & Valves for Nuclear Power APPENDIX C C.5-58 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

MAIN STEAM SAFETY VALVES (Continued)

Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required

7. Yoke Bending & Shear Sb = 18,200 lb/in2 Sb = 17,932 lb/in2 Stresses Sb = M , Ss = V Z A Ss = 10,900 lb/in2 Ss = 10,900 lb/in2 Loads:

Spring load at maximum where:

lift Bending & Shear Stress Sb = bending stress, lb/in2 Limits: Ss = shear stress, lb/in2 M = bending moment, in-lb Bending - allowable stress Z = section modulus, in3 intensity, Sm, per ASME V = vertical shear, lb Nuclear Pump & Valve Code A = shear area, in2 Shear - 0.6 x allowable stress intensity, 0.6 Sm, per ASME Nuclear Pump &

Valve Code.

Pd

8. Body Minimum Wall Thickness t = 15

. + C 2S 12

.P Loads: Body Bowl Primary Service pressure t = 0.562 in t = 0.3312 in Primary Stress Limit: where:

Inlet Nozzle Allowable stress, 7,000 P = primary service pressure, lb/in2, in accordance 150 lb/in2 t = 1.224 in t = 0.231 in with ASA B16.5. t = required thickness, in S = allowable stress, 7,000 lb/in2 Outlet Nozzle d = inside diameter of valve at section being considered, in t = 0.562 in t = 0.2823 in APPENDIX C C.5-59 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

MAIN STEAM SAFETY VALVES (Continued)

Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required

9. Inlet Nozzle Combined S = F1 + F2 + M1 + M2 S = 27,300 lb/in2 S = 5,997 lb/in2 Stress A Z Loads:

Spring load at maximum where:

lift Operational load S = combined bending & tensile Seismic load - maximum stress, lb/in2 credible earthquake F1= maximum spring load, lb F2= vertical component of reqction Combined Stress Limit: thrust, lb A = cross section area of nozzle, in2 1.5 x allowable stress M1= moment resulting from horizontal intensity, 1.5 Sm, per component of reaction, lb-in ASME Code for Pumps & M2= moment resulting from horizontal Valves for Nuclear Power. seismic force, in-lb

10. Spindle Diameter Actual Load Load Limit (0.2Fc)

Fc = 2EI Loads: L2 F = 110,383 lb F = 30,210 lb Spring load at maximum where:

lift Fc = critical buckling load, lb Spindle Column Load E = modulus of elasticity, lb/in2 Limit: I = moment of inertia, in4 L = length of spindle in compres-0.2 x critical buckling sion, in load APPENDIX C C.5-60 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

MAIN STEAM SAFETY VALVES (Continued)

Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required

11. Spring Washer Shear Ss = F Ss = 15,960 lb/in2 Ss = 2,312 lb/in2 Area A Loads: where:

Spring load at maximum Ss = shear stress, lb/in2 lift F = spring load, lb A = shear area, in2 Shear Stress Limit:

0.6 x allowable stress in-tensity, 0.6 Sm, per ASME Nuclear Pump & Valve Code.

APPENDIX C C.5-61 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

MAIN STEAM RELIEF VALVES Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required Pd

1. Body Minimum Wall Thickness t = 15

. + C Main Body:

. P 2S 12 Loads:

Design pressure & temperature t = 1.47 in t = 0.625 in Primary Membrane Stress Limit: where: Bonnet:

Allowable working stress t = minimum required thickness, in S = allowable stress, 7,000 lb/in2 t = 0.312 in t = 0.287 in as defined by USAS B16.5 P = primary service pressure, 655 (7,000 psi at primary d = inside diameter of valve at sec-service pressure). tion being considered, in C = corrosion allowance, 0.12 in

2. Bonnet Cap & Pilot Base Minimum Thickness CP 178

. WhG 1 2

t = d + + C1 Sm S m d3 Loads: Bonnet Cap:

Design pressure & temperature t = 1.0 in t = 0.612 in Gasket Load where: Pilot Base:

Primary Stress Limit:

t = minimum required thickness, in t = 2.219 in t = 2.117 in Allowable stress intensity, d = diameter or short span, in Sm, as defined by ASME Standard C = attachment factor, ASME Section Code for Pumps and Valves for VIII Nuclear Power. P = design pressure, lb/in2 Sm = allowable stress, lb/in2 W = total bolt load, lb hG = gasket moment arm, in C1 = corrosion allowance, 0.12 in APPENDIX C C.5-62 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

MAIN STEAM RELIEF VALVES (Continued)

Allowable Stress Calculated Stress or Criteria Method of Analysis or Design Dimension Minimum Dimension Required

3. Flange Bolt Area - Inlet Total bolting loads and stresses shall Body to Base:

Flange, Outlet Flange, Body be calculated in accordance with pro-to Bonnet, Bonnet to Base cedures of Para. 1-704.5.1 Flanged Ab = 10.26 in2 Ab = 2.854 in2 Joints, of B31.7 Nuclear Piping Code.

Loads: Bonnet to Cap:

Design pressure & temperature Ab = 1.452 in2 Ab = 0.995 in2 Gasket load Operational load Inlet Flange:

Maximum credible earthquake Ab = 13.9 in2 Ab = 6.25 in2 Bolting Stress Limit:

Outlet Flange:

Allowable stress intensity, Sm, as defined by ASME Ab = 12.2 in2 Ab = 5.5 in2 Standard Code for Pumps and Valves for Nuclear Power.

4. Flange Thickness - Inlet, Out- Flange thickness and stresses shall Body to Base:

let, Bonnet Flanges be calculated in accordance with pro-cedures of Paragraph 1-704.5.1 Flanged, SH = 26,250 lb/in2 SH = 24,412 lb/in2 Loads Joints of B31.7 Nuclear Piping Code. SR = 26,250 lb/in2 SR = 17,837 lb/in2 ST = 26,250 lb/in2 ST = 7,554 lb/in2 Design pressure & temperature Gasket load Cap to Bonnet:

Operational load Maximum credible earthquake SH = 26,250 lb/in2 SH = 15,598 lb/in2 SR = 26,250 lb/in2 SR = 3,325 lb/in2 Flanged Stress Limits, ST = 26,250 lb/in2 ST = 3,380 lb/in2 SH, SR, ST Inlet Flange:

1.5 Sm per ASME Nuclear Pumps and Valve Code. SH = 26,250 lb/in2 SH = 15,200 lb/in2 SR = 26,250 lb/in2 SR = 5,200 lb/in2 ST = 26,250 lb/in2 ST = 8,600 lb/in2 Outlet Flange:

SH = 26,250 lb/in2 SH = 12,437 lb/in2 SR = 26,250 lb/in2 SR = 12,213 lb/in2 ST = 26,250 lb/in2 ST = 3,088 lb/in2 APPENDIX C C.5-63 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

MAIN STEAM RELIEF VALVES (Continued)

Allowable Stress Calculated Stress or Criteria Method of Analysis or Design Dimension Minimum Dimension Required

5. Valve Disc. Thickness &

Stress Sr = St = 3 (3 + ) PR2 8 t2 Loads:

Design pressure & temperature where: Disc Stress:

Primary Stress Limit: Sr = radial stress, lb/in2 Sm = 15,800 lb/in2 St = Sr =

St = tangential stress, lb/in2 10,620 lb/in2 Allowable stress intensity, = Poisson's ratio Sm, as defined by ASME P = design pressure lb/in2 Standard Code for Pumps & R = radius of disc, in Valves for Nuclear Power. t = thickness of disc, in

6. Inlet Nozzle Diameter Thick- Inlet Nozzle ness & Stress S = F1 + F2 + M1 + M2 Stress:

A Z Loads: S = 26,250 lb/in2 S = 19,289 lb/in2 Design pressure & temperature where:

Operational load Maximum credible earthquake S = combined bending and tensile stress, lb/in2 Primary Stress Limit: F1 = vertical load due to design pressure, lb 1.5 x allowable stress F2 = vertical component of reaction intensity, 1.5 Sm as thrust, lb defined by ASME Standard A = cross section area of nozzle, in2 Code for Pumps & Valves for M1 = moment resulting from horizontal Nuclear Power reaction, in-lb M2 = moment resulting from horizontal seismic force at mass center of valve, in-lb APPENDIX C C.5-64 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

RECIRCULATION PUMPS Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required

1. Casing Minimum Wall t = PR + C 2.75 in 2.68 in Thickness SE - 0.6P Loads: Normal and where:

Upset Condition t = minimum required thickness, in Design pressure & P = design pressure, psig temperature R = maximum internal radius, in S = allowable working stress, psi Primary membrane stress E = joint efficiency limit: C = corrosion allowance, in Allowable working stress per ASME Section III, Class C APPENDIX C C.5-65 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

RECIRCULATION PUMPS (Continued)

Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required

2. Casing Cover Minimum Thickness b4(m 1) 4b4(m + 11 ) n a b + a2b2(m + 1) 3w 2 Sr = a 2b2 +

2 4t a (m 1) + b2(m + 1) 2 Loads: Normal and upset condition Design pressure &

temperature 2mb2 2b2(m + 11 ) n ab 3w

+ 1 = 15,075 psi Sr = 6.40 psi 2

2t a (m 1) + b (m + 1) 2 2 Primary Bending Stress Limit:

4 4 2 2 3w(m2 1) a b 4a b 1n a b 1.5 Sm per ASME code for St = +

2 2 2 4mt a (m 1) + b (m + 1)

Pumps and Valves for = 15,075 psi St = 5,243 psi Nuclear Power Class I ma2(m 1) mb2(m + 1) 2(m2 1)a21n a b 3w 1 +

2mt2 a2(m 1) + b2(m + 1) where:

Sr = radial stress at outer edge, psi St = tangential stress at inner edge, psi w = pressure load, psi W = uniform load along inner edge, lb t = disc thickness, in m = reciprocal of Poisson's ratio a - radius of disc, in b = radius of disc hole, in

3. Cover and seal flange Bolting loads, areas, and stresses Cover Flange Bolts bolt Areas shall be calculated in accordance 20,000 psi Loads: Normal and with "Rules for Bolted Flange 17,850 psi upset condition Connections" - ASME Section VIII, Appendix II. Seal Flange Bolts Design pressure & temperature Design gasket load 20,000 psi 17,750 psi Bolting Stress Limit:

APPENDIX C C.5-66 REV. 21, APRIL 2007

PBAPS UFSAR Allowable working stress per ASME Sect. III, Class C APPENDIX C C.5-67 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

RECIRCULATION PUMPS (Continued)

Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required

4. Cover Clamp Flange Flange thickness and stress shall be 9.25 in Flange Thickness &

Thickness calculated in accordance with "Rules Stress Loads: Normal and for Bolted Flange Connections" - ASME upset condition Section VIII, Appendix II. 8.9 in Design pressure & temperature Design gasket load Design bolting load Tangential Flange Stress Limit:

Allowable working stress per ASME Sect. III, Class C

5. Pump Nozzle Membrane SL = D2P + M + F Max. Pump Nozzle Stresses and Bending Stress 4 A Z A Suction Discharge Loads: Normal and upset condition SC = PD 2t Design pressure & temperature 28,650 psi S = 16,800 psi 16,800 psi Piping reactions during normal SS = TRO operation J Combined Stress Limit:

1 SL + SC S S 2 2 1.5 Sm per ASME code for Pumps S = L C 2

+ SS 2 2 and Valves for Nuclear Power Class I.

APPENDIX C C.5-68 REV. 21, APRIL 2007

PBAPS UFSAR APPENDIX C C.5-69 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

RECIRCULATION PUMPS (Continued)

Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required where:

SL = longitudinal stress, lb/in2 SC = circumferential stress, 2lb/in2 SS = torsional stress, lb/in2 D = nozzle internal diameter, in P = design pressure, lb/in2 A = nozzle cross section metal area, in M = maximum bending moment, in-lb F = maximum longitudinal force, lb t = nozzle wall thickness, in J = polar moment of inertia, in4 RO = nozzle outside radius, in T = torsional moment

6. Mounting Bracket Bracket vertical loads shall be deter- Maximum Combined Stress Combined Stress mined by summing the equipment and fluid weights and vertical seismic Bracket #1 #2&#3 Loads: forces. Bracket horizontal loads specific seismic force at mass center 17,300 psi 16,845 psi 11,458 psi Flooded weight shall be determined by applying the Maximum credible of pum-motor assembly (flooded).

earthquake Horizontal and vertical loads shall be applied simultaneously to determine Combined Stress Limit: tensile, shear and bending stresses in the brackets. Tensile, shear, and Yield Stress bending stress shall be combined to determine maximum combined stresses.

APPENDIX C C.5-70 REV. 21, APRIL 2007

PBAPS UFSAR APPENDIX C C.5-71 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

RECIRCULATION PUMPS (Continued)

Allowable Stress Calculated Stress or Criteria Method of Analysis or Actual Dimension Minimum Dimension Required

7. Stresses Due to Seismic The flooded pump-motor assembly shall Motor Bolt Tensile Stress:

Loads be analyzled as a free body supported by constant support hangers from the 11,200 psi 8,828 psi Loads: pump brackets. Horizontal and vertical seismic forces shall be applied at mass Pump Cover Bolt Tensile Operating pressure and center of assembly and equilibrium Stress:

temperature reactions shall be determined for the Maximum credible motor and pump brackets. Load, shear, 32,000 psi 19,148 psi earthquake and moment diagrams shall be construc-ted using live loads, dead loads, and Motor Support Barrel Combined Stress Limit: calculated snubber reactions. Combined Combined Stress:

bending, tension, and shear stresses Yield stress shall be determined for each major 22,400 psi 1,678 psi component of the assembly including motor, motor support barrel, boling, and pump casing. The maximum combined tensile stress in the cover boling shall be calculated using tensile stresses determined from loading diagram plus tensile stress from operating pressure.

APPENDIX C C.5-72 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

RECIRCULATION VALVES PB II Allowable Stress Calculated Stress or Criteria Method of Analysis orDesign Dimension Minimum Dimension Required 1.5 Pd

1. Body Minimum Wall In t = + 0.1 Pipe Run 2S-2P (1-y) 28" Suction Valve 28" Discharge Valve Loads:

Design Pressure where: 28" (Suction Valve) 28" (Suction Valve)

Design Temperature t = 1.940 in t = 1.667 in t = minimum wall thickness, in Codes and Standards P = design pressure, psig

1) USAS B31.1 1967 d = minimum diameter of flow passage,

2) Manufacturers Standards but not less than 90% of inside 28" (Discharge Valve) 28" (Discharge Valve)

Society MSS-SP.66 diameter at welding end, in t = 1.940 in t = 1.938 in S = allowable working stress, psi y = plastic stress distribution factor, 0.4

2. Body-to-Bonnet Bolt Total bolting loads and stresses Flange Bolt Area and Area Loads shall be calculated in accordance Stress with "Rules for Bolted Flange Sallow = 29,000 Am (in2) S (psi) lb/in2 2" Equal Bypass Valve Connections" except that the stem Ab (in2) 4" Discharge Bypass Valve operational load and seismic loads 22" Equalizer Valve carried by bolts. The horizontal 28" Suction Valve and vertical seismic forces shall 28" suc 33.46 14,838 28" Discharge Valve be applied at the equivalent mass 28" dis 39.13 17,634 center of the valve operator 28" suc 64.51 Loads: assuming that the valve body is 28" dis 64.51 rigid and anchored.

Design Pressure and Temp.

Gasket Load Stem Operational Load Seismic Load APPENDIX C C.5-73 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

RECIRCULATION VALVES (Continued)

Allowable Stress Calculated Stress or Criteria Method of Analysis or Design Dimension Minimum Dimension Required Bolting Stress Limit Allowable working stress per ASME Boiler & Pressure Vessel Code Sec. VIII App. II 1968 Edition Body Flange

3. Flange Stress Flange thickness and stress shall SH : 20,139 lb/in2 be calculated in accordance with (Hub Stress) SH SR ST 2" Equal. Bypass Valve "Rules for Bolted Flange Connections" SR : 13,426 lb/in2 4" Discharge Bypass Valve except that the stem thrust load and (Radial Stress) 22" Equalizer Valve seismic loads shall be included in ST : 13,426 lb/in2 28" Suction Valve the total load carried by the flange. (Tangential 28" Discharge Valve The horizontal and vertical seismic Stress) 28"suc 14,474 9,111 6,113 forces shall be applied at the 28"dis 15,966 10,265 6,772 equivalent mass center of the valve operator assuming that the valve body is rigid.

Loads:

Bonnet Flange Design pressure & SH SR ST temperature Gasket Load Stem Operational Load Seismic Load 28"suc 13,160 10,969 6,826 Codes--ASME Boiler & 28"dic 14,211 12,308 7,930 Pressure Vessel Code Section VIII Appendix II, 1968 Flange Stress Limits; S , S , S :

S per ASME Boiler &

Pressure Vessel Code Sec. VIII App. II, 1968 Edition APPENDIX C C.5-74 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

RECIRCULATION VALVES (Continued)

Allowable Stress Calculated Stress or Criteria Method of Analysis or Design Dimension Minimum Dimension Required

4. Yoke Bolts The valve assembly is analyzed Sb (allowable) = Sb, (bolt stress, Sb assuming that the valve body is an 20,000 lb/in2 lb/in2) anchored, rigid mass and that the specified vertical and horizontal seismic forces are applied at the equivalent mass center of the 28" Suction Valve operator assembly equivalent.

28" Discharge Valve Using these forces, stresses and 28" (Suction) 1,322 deflections are determined for the 28" (Discharge) 6,326 operator support components.

Loads:

Seismic load APPENDIX C C.5-75 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

RECIRCULATION VALVES PB III Allowable Stress Calculated Stress or Criteria Method of Analysis or Design Dimension Minimum Dimension Required

1. Body Minimum Wall In 2" (Equal. 2" (Equal.

Pipe Run t = 1.5 Pd + 0.1 Bypass Valve) Bypass Valve) 2S-2P (1-y) t = 0.400 in t = 0.253 in 2" Equal. Bypass Valve 4" (Disch. 4" (Disch.

4" Discharge Bypass Valve Bypass Valve) Bypass Valve) 22" Equalizer Valve t = 0.700 in t = 0.405 in 28" Suction Valve 28" Discharge Valve 22" (Equal. Valve) 22" (Equal. Valve) t = 1.540 in t = 1.520 in Loads: where: 28" (Suction Valve) 28" (Suction Valve) t = 1.940 in t = 1.677 in Design Pressure t = minimum wall thickness, in Design Temperature P = design pressure, psig 28" (Discharge Valve) 28" (Discharge Valve) d = minimum diameter of flow passage, t = 1.940 in t = 1.938 in Codes and Standards but not less than 90% of inside

1) USAS B31.1 1967 diameter at welding end, in

2) Manufacturers Standards S = Allowable working stress, psi Society MSS-SP.66 y = plastic stress distribution factor, 0.4

2. Body-to-Bonnet Bolt Flange Bolt Area and Area Loads Stress 2" Equal Bypass Valve Total bolting loads and stresses Sallow = 29,000 psi 4" Discharge Bypass Valve shall be calculated in accordance Ab (in2) Sb (psi) with "Rules for Bolted Flange Ab (in2) 2" 1.68 26,883 Loads: Connections" except that the stem 2" 1.81 4" 3.40 10,613 operational load and seismic loads 4" 9.29 Design pressure & tempera- shall be included in the total load ture carried by bolts. The horizontal Gasket Load and vertical seismic forces shall Stem Operational be applied at the mass center of Load the valve operator assuming that Seismic Load the valve body is rigid and anchored.

APPENDIX C C.5-76 REV. 21, APRIL 2007

PBAPS UFSAR APPENDIX C C.5-77 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

RECIRCULATION VALVES (Continued)

PB III Allowable Stress Calculated Stress or Criteria Method of Analysis or Design Dimension Minimum Dimension Required Bolting Stress Limit Allowable working stress per ASME Boiler & Pressure Vessel Code Sec. VIII App. II 1968 Edition Body Flange

3. Flange Stress Flange thickness and stress shall SH: 20,139 lb/in2 be calculated in accordance with (Hub Stress) SH(psi) SR(psi) ST(psi) 2" Equal. Bypass Valve "Rules for Bolted Flange Connections" SR: 13,426 lb/in2 4" Discharge Bypass Valve except that the stem thrust load and (Radial Stress) 2" 10,175 5,103 9,352 seismic loads shall be included in ST: 13,426 lb/in2 4" 13,408 6,303 11,935 the total load carried by the flange. (Tangential Stress)

Loads: The horizontal and vertical seismic Bonnet Flange forces shall be applied at the mass Design pressure & temperature center of the valve operator assuming Gasket load that the valve body is rigid. SH SR ST Stem Operational load Seismic Loads 2" NA NA NA 4" NA NA NA Codes--ASME Boiler &

Pressure Vessel Code Section VIII App. II ASME Boiler & Pressure Vessel Code,Section VIII, App. II, 1968 Edition.

APPENDIX C C.5-78 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

RECIRCULATION VALVES (Continued)

PB III Allowable Stress Calculated Stress or Criteria Method of Analysis or DesignDimension Minimum Dimension Required

4. A) Body & Bonnet Flange Primary, secondary, and peak stresses Primary Membrane 22" Valve 28" Valve Stress were analyzed using finite element plus bending =

B) Body Neck Wall Stress computer analyses. The model was 23,700 psi 20,660 psi 22,770 psi 22" Equalizer Valves verified by strain gage tests. The 28" Suction Valves seismic and stem thrust loads are to Local Membrane = 16,168 psi 17,860 psi 28" Discharge Valves be converted to an equivalent pres- 23,700 psi sure and proportional stress added directly. Secondary Membrane 98,230 psi* 48,750 psi*

plus Loads: Bending Range = *See ASME Code Case 47,400 psi No. 1441.

Design pressure &

temperature Seismic Load Stem Thrust Codes--ASME Boiler &

Pressure Vessel Code Sec. III APPENDIX C C.5-79 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.8 (Continued)

RECIRCULATION VALVES (Continued)

PB III Allowable Stress Calculated Stress or Criteria Method of Analysis or Design Dimension Minimum Dimension Required 22" Valve 28" Suc.& Disch.

5. Body-to-Bonnet Bolting Primary, secondary, and peak stresses Under operating 33,900 psi 42,900 psi were analyzed using finite element conditions 67,000 psi 22" Equalizer Valve computer analyses. The model was 28" Suction & Discharge verified by strain gage tests. The Maximum conditions 85,000 psi 59,000 psi Valve Seismic and stem thrust loads are to 100,500 psi be converted to an equivalent Loads: pressure and proportional stress added directly.

Design Pressure Design Temperature Seismic Load Stem Thrust Codes--ASME Boiler &

Pressure Vessel Code Sec. III 1968 Edition

6. Valve Operator Support The valve assembly is analyzed Sb allowable = Sb, (bolt stress, psi) Sb Bolting assuming that the valve body is an 20,000 psi anchored, rigid mass and that the 2" (Equal. Bypass) 3,532 2" Equal. Bypass Valve specified vertical and horizontal 4" (Discharge Bypass) 10,622 4" Discharge Bypass Valve seismic forces are applied at the 22" (Equal.) 2,602 22" Equalizer Valve mass center of the operator assembly. 28" (Suction) 2,906 28" Suction Valve Using these forces, stresses and 28" (Discharge) 3,840 28" Discharge Valve deflections are determined for the yoke leg to bonnet bolts.

Loads:

Equipment dead weight Seismic load -

Codes--ASME Boiler &

Pressure Vessel Code Section VIII 1968 Edition.

APPENDIX C C.5-80 REV. 21, APRIL 2007

PBAPS UFSAR TABLE C.5.9 DAMPING FACTORS FOR MARK I LOADS Percent of Critical Damping Piping Diameter NPS 12 inch NPS 12 inch Safety relief valve 1 2 discharge loading Pool swell, condensation oscillation, chugging 2 3 APPENDIX C C.5-81 REV. 21, APRIL 2007